Text

License Renewal Application Amendment 4: Response to Request for Additional Information Regarding Severe Accident Mitigation Alternatives
	ML061930418
Person / Time
Site:	Pilgrim
Issue date:	07/05/2006
From:	Bethay S; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	%dam200612, 2.06.058, TAC MC9676
	Download: ML061930418 (71)
	v • d • e

AE y

'"ýEnteigy Entergy Nuclear Operations, Inc.

Pilgrim Station 600 Rocky Hill Road Plymouth, MA 02360 July 5, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Stephen J. Bethay Director, Nuclear Assessment

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No.: 50-293 License No.: DPR-35

REFERENCES:

License Renewal Application Amendment 4:

Response to Request for Additional Information Regarding Severe Accident Mitigation Alternatives for Pilgrim Nuclear Power Station (TAC NO. MC9676)

1. Entergy letter, License Renewal Application, dated January 25, 2006 (2.06.003)

2. NRC letter, Request for Additional Information, dated May 22, 2006 2.06.058 LETTER NUMBER:

Dear Sir or Madam:

In Reference 1, Entergy applied for the renewal of the Pilgrim Station operating license. The application included Appendix E, Applicant's Environmental Report. In Reference 2, the NRC requested additional information regarding severe accident mitigation alternatives (SAMAs) submitted in Reference 1 (Appendix E).

The attachment to this letter provides the additional information requested in Reference 2.

By this letter Entergy also notes a typographical error in Section E.1.5.2.9 (page E.1-66) of Attachment E.1, Evaluation of PSA Model, of the Environmental Report submitted in Reference

1. Specifically, 'Twelve release categories... " should state "Nineteen release categories..

This letter contains no new commitments.

Please contact Mr. Bryan Ford, at 508-830-8403, if you have any questions regarding this subject.

I declare under the penalty of perjury that the foregoing is true and correct. Executed on the 5-L-"

day of July 2006.

DWE/dm

Attachment:

Response to Request for Additional Information Regarding SAMAs Afin W1 9

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Letter Number: 2.06.058 Page 2 cc: with Attachment Mr. Alicia Williamson Project Manager U.S. Nuclear Regulatory Commission Office 0-11 F1 11555 Rockville Pike Rockville, MD 20852 cc: without Attachment Mr. James Shea U.S. Nuclear Regulatory Commission Office O-8B-1 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Jack Strosnider, Director Office of Nuclear Material and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-00001 Mr. Samuel J. Collins, Administrator Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Joseph Rogers Commonwealth of Massachusetts Assistant Attorney General Division Chief, Utilities Division 1 Ashburton Place Boston, MA 02108 Mr. Robert Walker, Director Massachusetts Department of Public Health Radiation Control Program Schrafft Center, Suite 1 M2A 529 Main Street Charlestown, MA 02129 Ms. Cristine McCombs, Director Massachusetts Emergency Management Agency 400 Worchester Road Framingham, MA 01702 Mr. James E. Dyer, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-00001 NRC Resident Inspector Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360

ATTACHMENT A to Letter 2.06.058 Response to Request for Additional Information Regarding SAMAs

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE ANALYSIS OF SEVERE ACCIDENT MITIGATION ALTERNATIVES (SAMAs)

FOR THE PILGRIM NUCLEAR POWER STATION (PNPS)

DOCKET NO. 50-293 Table of Contents NRC RAI 1.......................................................................................................................................

3 Response to RAI la.....................................................................................................................

3 Response to RAI lb.....................................................................................................................

4 Response to RAI lc.............................................................................................................

4 Response to RAI Id.....................................................................................................................

4 Response to RAI le.....................................................................................................................

4 Response to RAI If......................................................................................................................

5 NRC RAI 2.......................................................................................................................................

6 Response to RAI 2a.............................................................................................................

7 Response to RAI 2b.............................................................................................................

8 Response to RAI 2c.............................................................................................................

9 Response to RAI 2d...................................................................................................................

10 Response to RAI 2e...................................................................................................................

10 NRC RAI 3.....................................................................................................................................

29 Response to RAI 3a...................................................................................................................

29 Response to RAI 3b...................................................................................................................

30 Response to RAI 3c...................................................................................................................

31 Response to RAI 3d...................................................................................................................

32 NRC RAI 4.....................................................................................................................................

41 Response to RAI 4a..................................................................................................................

41 Response to RAI 4b...................................................................................................................

41 Response to RAI 4c...................................................................................................................

41 NRC RAI 5.....................................................................................................................................

42 Response to RAI 5a...................................................................................................................

43 Response to RAI 5b.....................................................

............................................................. 43 Response to RAI 5c...................................................................................................................

44 Response to RAI 5d...................................................................................................................

44 Response to RAI 5e...........................................................................................................

45 Response to RAI 5f....................................................................................................................

45 Response to RAI 5g...................................................................................................................

46 Response to RAI 5h...................................................................................................................

46 1 of 68

NRC RAI 6......................................................................................................................................

59 Response to RAI 6a...................................................................................................................

59 Response to RAI 6b...................................................................................................................

61 Response to RAI 6c...................................................................................................................

62 Response to RAI 6d...................................................................................................................

62 Response to RAI 6e...................................................................................................................

62 Response to RAI 6f....................................................................................................................

62 Response to RAI 6g...................................................................................................................

63 NRC RAI 7.....................................................................................................................................

67 Response to RAI 7a...................................................................................................................

67 Response to RAI 7b...................................................................................................................

67 Response to RAI 7c...................................................................................................................

68 Response to RAI 7d...................................................................................................................

68 Response to RAI 7e...................................................................................................................

68 Response to RAI 7f....................................................................................................................

68 Figures Figure RAI.2-1 Level I-to-Level II Interface Logic Tree...............................................................

12 Figure RAI.5-1 Fire W ater Cross-Tie to LPCI.............................................................................

57 Figure RAI.5-2 PNPS Direct Torus Vent Pathway.......................................................................

58 Tables Table RAI.2-1 Table RAI.2-2 Table RAI.2-3 Table RAI.2-4 Table RAI.3-1 Table RAI.3-2 Table RAI.5-1 Table RAI.6-1 "Rebinned" Plant Damage States........................................................................

13 Summary of PNPS Core Damage Accident Sequences Plant Damage States....... 14 Grouping of Source Term Magnitude....................................................................

19 Description of PNPS CET Release Modes...........................................................

20 PNPS Dominant Scenarios to Fire CDF

.................................. 33 Revised Summary of Phase II SAMA Analysis.....................................................

35 Phase I SAMA Analysis (SAMA 248 through 281)..............................................

47 Reduction in Off-site Economic Cost Risk (OECR)..............................................

64 2 of 68

NRC RAI 1 The SAMA analysis is said to be based on the most recent version of the PNPS Probabilistic Safety Analysis (PSA) (Revision 1 April 2003). Provide the following information regarding these PSA models:

a. The PNPS individual plant examination (IPE) evaluated total and partial loss of offsite power events. The current PSA model includes only a single loss of offsite power (LOOP) event.

Characterize this LOOP event relative to the IPE events.

b. It is stated that the PSA represents the plant operating configuration and design changes as of September 30, 2001. Identify any changes to the plant (physical and procedural modifications) since September 2001 that could have a significant impact on the results of the PSA. Provide a qualitative assessment of their impact on the PSA and their potential impact on the results of the SAMA evaluation.

c. The Boiling Water Reactor Owners Group (BWROG) peer review in 2000 apparently reviewed the original 1992 IPE instead of the 1995 revision. Explain why the 1995 revision was not peer reviewed.

d. The environmental report (ER) states that all major issues and observations from the BWROG peer review have been addressed and incorporated in the current PSA. Describe the "non-major' issues that have not been incorporated and their potential impact on the results of the SAMA evaluation. Discuss the overall conclusion of the BWROG peer review relative to the use of the Pilgrim PSA.

e. The description of the revisions of the peer reviewed 1992 IPE to produce the current 2003 PSA indicates that almost all of the elements of IPE were completely revised. Provide more detail on the steps taken to ensure the technical adequacy of the current Level 1 and Level 2 PSA, including the review criteria used, a summary of the results of the peer review described in paragraphs 2 and 4 of ER Section E.1.4.1, and an identification of any open items from this review and their potential impact on the conclusions of the SAMA analysis.

f.

The ER appears to provide a listing of the major plant and PSA model changes since the 1995 IPE Update. However, it is not clear whether these changes include differences between the 1992 IPE and 1995 IPE update. Provide a listing of the changes between the 1992 and 1995 models and between the 1995 and 2003 models. Indicate which changes were the major contributors to the reduction in core damage frequency (CDF).

Response to RAI 1 a

a. The IPE submittal modeled two frequencies for loss of offsite power. One was the entire loss of 345kV and 23kV together. The other was the partial loss of 345kV with 23kV available. This partial loss was derived as a fraction of the losses of 345kV and 23kV.

The 2003 PSA model used one single frequency for loss of offsite power from the 345kV ring bus. Loss of the 23kV feed from the Manomet Station to the shutdown transformer was modeled as a split fraction (i.e. conditional probability) of this frequency. It was conservatively assumed that 50% of the losses of offsite power resulted in a complete loss of all incoming AC power, despite the independence of the 23kV line.

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Response to RAI lb

b. There have been no plant changes that could have a significant impact on the results of the SAMA analysis since the model freeze date of September 2001.

Plant changes since the freeze date with potential minor impact on the PSA results are discussed below.

A new 100% capacity self contained diesel driven rotary screw IAS compressor, K-1 17, replaced the function of the smaller reciprocating compressors, K-1 04A, K-1 04B, and K-104C which have been placed in wet layup. This plant modification results in no appreciable change in CDF.

Plant procedure 5.3.26, "RCIC operation without DC power," provides operational guidance to locally control RCIC operation following a catastrophic event in which both AC and DC power become unavailable and a reactor pressure vessel injection system is required to restore and maintain reactor water level. The impact of this procedural change will be evaluated during the next PSA update, but would reduce the CDF impact of AC and DC power failures. Since they impact the same sequences, SAMAs related to AC and DC power failures may have a slightly reduced benefit if this change was in the model used for the SAMA analysis. Therefore, the conclusions of the SAMA analysis are conservative with relation to modeling of this change.

Operator action to manually initiate HPCI/RCIC for sequences that involve auto initiation signal failure has been proceduralized. Since CDF is dominated by loss of containment decay heat removal sequences, and manual start of HPCI/RCIC does not impact these sequences, this change has no appreciable change on CDF.

Since these changes have minor potential impact on CDF, the conclusions of the SAMA analysis would be unchanged if they were included in the model used for the SAMA analysis.

Response to RAI 1c

c. The Boiling Water Reactor Owners Group (BWROG) peer review in 2000 reviewed the original 1992 IPE as well as the minor changes incorporated in the 1995 revision. Thus, the 1995 revision was included in the BWROG Peer review.

Response to RAI Ild

d. The BWROG Peer Review concluded the Pilgrim IPE can be effectively used to support applications when significant issues are addressed. All issues identified by the BWROG peer review have been incorporated into the 2003 PSA model. In addition, an independent assessment team from Entergy South concluded that the 2003 model met the pertinent aspects of NEI-00-02, "Probabilistic Risk Assessment Peer Review Process Guidance."

Therefore, this model is appropriate for use in the SAMA analysis.

Response to RAI l e

e. Several steps were taken to assure the technical adequacy of the 2003 PSA model.

Individual work packages (event tree, fault tree, human reliability analysis (HRA), data, etc.)

and internal flooding analysis were circulated to each PSA member for independent peer 4 of 68

review. The accident sequence packages, system work packages, HRA, and internal flooding analyses were also assigned to the appropriate plant personnel for review. For example, event trees, system analyses, and fault tree models were forwarded to the applicable plant systems engineers and the HRA was assigned to individuals from the plant Operations Training department for review. Prior to issuance, the 2003 PSA model was reviewed externally through the use of an independent team of consultants.

The independent team of consultants concluded that the PRA revision had been performed in a logical, reasonable, and thorough manner and that although certain changes were recommended, none of these changes would require a major revision of the analysis or the results obtained. Recommended changes were examined with the review team and appropriate changes were made to the analysis and the report. Implementation of the remaining changes would not impact the conclusions of the SAMA analysis.

Subsequent to issuance of the 2003 PSA model, an independent team of PSA analysts from Entergy South reviewed the model against NEI-00-02 "Probabilistic Risk Assessment Peer Review Process Guidance." The team concluded that the 2003 PSA model addressed the appropriate elements. Additionally, they found that the update process was implemented in a manner that properly documents the model and supporting analysis. The update and maintenance process of the PSA model are conducted in accordance with the provisions of NEI-00-02. Therefore, the 2003 PSA model is appropriate for use in the SAMA analysis.

Response to RAI If

f. In 1995, the original IPE model was changed in response to the NRC Request for Additional Information (RAI) received in April 1995. Overall CDF was reduced from 5.85E-5/yr to 2.84E-5/yr. The reduction in CDF was due to removal of HPCI room cooling dependency, revised ADS success criteria, and improved HPCI/RCIC performance.

The 2003 PSA model is based upon all procedures and plant design as of September 30, 2001, and plant data as of December 31, 2001. The results yield a measurably lower CDF (point estimate CDF = 6.41 E-6/reactor year) than the 1995 PSA model update (point estimate CDF = 2.84E-5/yr). The improved results since the 1995 model are due to improved plant performance, replacement of switchyard breakers, more realistic success criteria based on MAAP runs, and more sophisticated data handling. Major changes are summarized in ER Section E.1.4.2.

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NRC RAI 2 Provide the following information relative to the Level 2 analysis:

a. In ER Section E.1.2.2.1 it is stated that 'The Lbvel 1 and plant system information is passed through to the [containment event tree] (CET) evaluation in discrete (plant damage state]

(PDS)." ER Table E.1-4 identifies seven PDS groups and ER Table E.1-8 identifies 48 more detailed PDSs. It is noted that for certain PDSs, the frequency in the Table E.1-4 does not equal the sum of the frequencies for like-PDSs in Table E.1-8. Provide a description of the mapping of Level 1 results into the various containment end states/release categories, and the relevance of the PDS as input to the CET. Address whether the PDSs uniquely define failed equipment for the CET analysis or whether this is done by inputting the cutsets. Also, discuss whether the sequences that make up a PDS are combined and entered into the CET as a frequency, or whether the cutsets that make up each group of core damage sequences are entered into the CET, and the relevance of the two inconsistent sets of frequency values in Tables E.1-4 and E.1-8.

b. ER Table E.1-7 defines 7 release categories and Table E.1-10 provides the frequency of these categories. Source term characteristics are, however, defined for 19 collapsed accident progression bins (CAPBs) in Tables E.1 -9 and E.1 -11. There appears to be some disconnect between the release categories and the CAPBs. For example, CAPB-15 is indicated to involve late containment failure (Table E.1-9) and a high Csl release fraction of 27 percent (Table E.1-11), yet Table E.1-10 indicates the frequency of Late High release is 0.0. Also, none of the so-called late containment failure CAPBs have release start times greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (8.64E+04 seconds) which is Entergy's definition of late. Describe the use of the release categories and how they are related to the CAPBs.

c. With regard to source terms, provide the following information:

i. Briefly describe the approach used to determine the source terms for each release category. Clarify whether new MAAP analyses were performed as part of the development of the current model and how the MAAP cases were selected to represent each release category (i.e., based on the frequency dominant sequence in each category or on a conservative, bounding sequence).

ii. ER Section E.1.2.2.6 indicates that the source terms were grouped into a much smaller number of source term groups with frequency-weighted mean source terms for each group. Clarify whether the source terms prior to this grouping process correspond to the accident sequence-CET endpoints, and the smaller number of source term groups correspond to the CAPBs. Discuss the development of a frequency-weighted mean source term for each group.

d. ER Section E.1.2.2.6 indicates that the accident progression bins for each of the 48 PDS were sorted into the CAPBs based on a number of attributes. Not included in the list are the CET fission product removal and reactor building nodes identified in Table E.1-5 or containment venting. These would appear to impact the release fractions. Please explain.

e. Only about 3 percent of the CDF leads to early containment failure, with the majority of the releases occurring late (after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following event initiation). Explain this relatively small percentage in terms of the early containment failure modes associated with Mark I containments, including liner melt-through by molten core debris and containment venting.

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Clarify how sequences involving containment venting (from the suppression chamber or the drywell) are assigned using the release categories of ER Table E.1-10.

Response to RAI 2a

a. Specific plant damage state (PDS) binning used a two-stage process. First, eleven questions were developed to properly describe the state of the reactor, containment and core cooling systems as the accident proceeds to core damage. These questions are:

(1) What is the initiating event?

(2)

Is the containment bypassed?

(3)

Is offsite power available?

(4)

Is onsite power available?

(5)

Are high-pressure systems (HPCI, RCIC, CRD, or SLC) available?

(6)

Are low-pressure systems (LPCI, core spray, condensate, firewater) available?

(7)

What is the status of reactor pressure?

(8)

Is RHR available for decay heat removal?

(9) Are drywell sprays available for ex-vessel injection?

(10) Is the containment vented before core damage?

(11) When does core damage occur?

While the total number of PDSs would be very large if all combinations of answers to the above questions represented unique PDSs, some combinations are illogical and thus are eliminated. Other combinations can be eliminated because they will not result in different accident progressions and therefore, different containment failure and source term characteristics.

The binning criteria, based on the above questions were reduced to the following PDS plant-specific attributes:

Type of initiating event (question 1)

Containment bypass (question 2)

Availability of AC power (questions 3 and 4)

RCS pressure during the accident sequence and at vessel breach (question 7)

Time to core damage (question 11)

The provision of high-pressure vessel make-up during the course of the accident (question 5)

" The provision of low-pressure vessel make-up during the course of the accident (question 6)

" The ability to remove heat from the containment atmosphere (questions 8, 9, and 10)

The refined criteria reflect important distinctions between initiating events, (i.e., transient versus LOCAs) system functional attributes (i.e., the availability of containment heat removal), and system operation (i.e., high or low RCS pressure).

Based on these criteria, a Level 1-to-Level 2 interface event tree was developed. Figure RAI.2-1 depicts the selected PDS binning attributes. This event tree also serves as the template for sorting individual Level 1 sequences into the appropriate plant damage states.

Each sequence is characterized by a set of conditions which define a unique plant damage state.

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The PDSs, based on Level 1 sequences uniquely define the status of the reactor (high pressure or low), containment (intact or bypass), and core cooling systems (is high-pressure injection from HPIC/RCIC failed or is low pressure injection from LPCI, core spray, etc.

available) at the time of core damage.

The PDS frequency is represented by the total frequency of those Level 1 sequences that satisfy the PDS binning criteria shown in Figure RAI.2-1. For example, from the updated PSA, sequence A-3 is a large LOCA sequence that results in core damage due to failure of low-pressure injection systems LPCI and core spray. This sequence satisfies PDS-7 attributes: LOCA initiator, low vessel pressure, early core damage, and the availability of alternate low-pressure injection systems (firewater), containment decay heat removal (torus cooling, etc.) and torus venting.

The input to the containment event tree (CET) is the PDS frequency and different flag settings based on the PDS definition as depicted in Figure RAI.2-1. Hence, the CET outcome represents the unique accident progression for that PDS.

The enclosed Table RAI.2-1 presents the binning scheme used to generate the plant damage state groups provided in Table E.1-4 of the ER. As stated in Section E.1.2.2.5, the Level 1-to-Level 2 binning process defined 48 PDSs. However, to facilitate the description of the Level 2 analysis, the PDSs were re-examined to determine if PDSs that reflect similar reactor and containment states could be combined or "rebinned". The method used was to reclassify the PDS initiating events into general initiator groups (LOCAs, transients, SBO, vessel rupture, ATWS, ISLOCA and transient with loss of containment decay heat removal

[TW]). PDSs that exhibit similar reactor coolant system and containment states were then combined if the initiating events belonged to the same initiator group. Table RAI.2-1 lists the association of the general initiator groups. By rebinning, the number of PDSs was reduced from 48 to 7. However, it must be stressed that all 48 PDSs were addressed in the containment event tree; this plant damage state group rebinning is for presentation or descriptive purposes only.

In addition, upon reviewing the contents of Table E.1-8, for PDS-5 the correct value should have been 5.59 x 1010/ry instead of 0.0. As a result, Table RAI.2-2 is issued as an update of Table E.1-8.

Response to RAI 2b

b. The following information replaces the discussion in Sections E.1.2.2.2 through E.1.2.2.4, Table E.1-6, Table E.1-7, and the "Nomenclature" portion of Table E.1-10 of the ER. (The discussion for another plant was inadvertently placed in the ER.)

The release category magnitude bin assignment used in the SAMA evaluation is based on cesium iodine (Csi) and tellurium (Te) release fractions alone. The CsI release fraction indicates the fraction of in-vessel radionuclides escaping to the environment. The tellurium release fraction indicates the fraction of products of core-concrete interactions that escape.

Noble gas releases are considered essentially complete given containment failure. Table RAI.2-3 indicates the scheme used to make this assignment. Based on this release magnitude bin assignment, no late high release category results.

The CET is used as the starting point to bin different accident progressions endstates into specific release categories. Each CET end state represents a particular release event or a 8 of 68

recovered, degraded core state that may be characterized according to its potential for fission product releases to the atmosphere, its timing of release initiation relative to time of incipient core damage, and its release duration.

Table RAI.2-4 summarizes the possible containment event tree release endstates for the spectrum of core melt accident sequences and the associated collapsed accident progression bin (CAPB). This table defines the various containment event tree release modes in terms of the occurrence of core damage, the occurrence of vessel breach, primary system pressure at vessel breach, the location of containment failure, the timing of containment failure and the occurrence of core-concrete interactions.

Release timings for the CAPBs were assigned by analogy with the source terms reported for Peach Bottom, Unit 2, NUREG/CR-4551 Volume 4, Revision 1, Part 2. The information contain on Page B.2-1 of NUREG/CR-4551 were judged, based on the descriptions, to be similar in character to the CAPB release modes. Hence, the assigned release timings for early release CAPBs are based on the following:

3.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or 12960 seconds (Fast ATWS, Fast SBO, LOCA: failure near VB) 6.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or 21960 seconds (Fast TC, Fast SBO, LOCA: late failure) 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months or 27000 seconds (Slow SBO: failure near VB)

Similarly the assigned release timings for late release CAPBs are based on either 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months s/36000 seconds (Slow SBO, late failure) or 13 hour1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months s/46800 seconds (Very Slow SBO; late failure).

Response to RAI 2c C.

i. The magnitude of the source term release resulting from CET accident progression was estimated using a source term algorithm. This algorithm is a set of algebraic expressions that calculate release of each radionuclide group to the environment based on the release from fuel debris and removal mechanisms active in the severe accident progression.

The basic parametric equation used in calculating the source term magnitude is:

Ren(i)= R-R(i) + R ca(i) + R REV(i)

Where:

The first term RIv (i) represent the releases to the environment due to core melt in-vessel.

The second term Rcc, (1) represents ex-vessel releases from core-concrete interactions that is released to the environment.

The third term RREV (1) represents releases to the environment due to revolatilization release from the primary coolant system after vessel breach (I, CS and TE only)

The above individual releases to the environment are a function of the individual fraction of the available material in a given fission product group that evolves from the core 9 of 68

debris and becomes available for release to the environment, divided by the deposition mechanisms that act on this material to limit its ultimate release to the environment.

The use of the above source term algorithm and MAAP calculations generates the source term estimates used in characterizing the severity of the containment event tree endstates. The source term estimates in turn are used in the MACCS2 consequence analysis. MAAP is an industry recognized thermal hydraulics code used to evaluate design basis and beyond design basis accidents. MAAP (Version 4.04) and MAAP parameter file (pnps4-sa.par) have been used in this evaluation. The parameter file contains plant specific parameters representing the primary system and containment.

The 2003 PSA model documents MAAP calculations which are representative deterministic thermal hydraulic calculations that portray dominant CET scenarios.

ii. The CAPBs source terms described in Section E.1.2.2.6 of the ER and used in the consequence analysis are generated by sorting all of the CET accident progression bins for each plant damage state (see Table RAI.2-2) on attributes of the accident progression. These collapsed bins are composed essentially of six characteristics: the occurrence of core damage, the occurrence of vessel breach, primary system pressure at vessel breach, the location of containment failure, the timing of containment failure and the occurrence of core-concrete interactions.

The CAPBs source terms are represented by frequency weighted mean source terms.

This process entailed the following steps:

1. Determine the mean frequency of each CAPB by summing the individual mean PDSs accident progressions CET endpoint frequencies contained in the particular CAPB.

2. Determine the CAPB individual conditional probability for each CET accident progression by dividing the result from Step 1 into the individual PDSs frequencies.

3. Multiply each PDS accident progression CET endpoint source terms, release timing, release energy and release elevation by the value determine in Step 2.

4. Sum the individual results of Step 3 to arrive at the total final values contained in Table E.1-11 of the ER.

Response to RAI 2d

d. The impact of fission product removal (i.e., drywell sprays), reactor building fission product retention and torus pool scrubbing (which indirectly includes torus venting) is accounted for when estimating the source terms associated with a particular CET endstate.

Response to RAI 2e

e. Transients with loss of long-term containment decay heat removal (TW sequences) dominate the internal CDF, representing 91.5% of the total CDF. TW sequences entail loss of the torus cooling and drywell spray modes of RHR. The loss of containment heat removal results in elevated containment pressures and eventual containment failure. Containment release results from either direct torus venting or steam overpressurization failure. Because TW sequences result in late containment failure, early containment failure, as a percentage of CDF is correspondingly low. Early containment failures are dominated by SBO and 10 of 68

ATWS sequences with the dominant containment failure modes being drywell liner melt-through, drywell/torus overpressurization at vessel breach and reactor pedestal overpressurization at vessel breach. These results are consistent with results for other MARK I containments.

For sequences involving containment venting and subsequent core damage (PDSs 1, 5, 12, 18, 40 and 43), the impact of containment venting on the release category is considered directly in the Level 2 fault tree developed for fission product removal. For these plant damage states, the occurrence of successful venting, given a TW event, results in flag

'FLAG-VENT-OK' being set to true. As a result, the likelihood for successful fission product removal increases. Hence, CET accident progressions that involve successful fission product removal results in lower release category when compared to CET accident progressions in which no fission product removal occurs.

11 of 68

Figure RAI.2-1 Level I-to-Level II Interface Logic Tree CORE ODAJAE WAT IS THE REV PRSSURE D

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Table RAI.2-1 "Rebinned" Plant Damage States Point

% of Total Core PDS Group Plant Damage States Esimt Damag Fre Estimate Damage Frequency Loss of Coolant Accidents 3, 4, 7, 8, 9, 10, 11 and 37 1.16 x 10.7 1.80 Transients 14, 15, 16, 17, 20, 21, 22, 23, 2.43 x 10.7 3.79 24, 25, 26, 27, 28, 38, 39, 41,42 Station Blackout 29, 30, 31 and 32 1.48 x 10-7 2.31 Vessel Rupture 33, 34, 35 and 36 4.00 x 10-0.06 Anticipated Transients 45 and 46 3.39 x 108 0.53 without Scram Loss of Containment Heat 1,2,5,6,12,13,18,19,40,43 and 5.86 x 91.45 Removal ('TW')

44 Inter-System LOCA 47 and 48 4.00 x 10-9 0.06 13 of 68

Table RAI.2-2 Summary of PNPS Core Damage Accident Sequences Plant Damage States Point

% OF PDS Description Estimate CDF Long-term LOCA with loss of high-pressure core makeup from HPCI and RCIC, loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core damage results at high primary system 0.00E+00 0.0 PDS-1 pressure. Late injection from low-pressure systems (core spray, LPCI, and firewater) is available, provided primary system depressurization occurs. The containment is vented and intact.

Long-term LOCA with loss of both high-pressure core makeup (HPCI and RCIC) and containment heat removal. Core damage results at high primary system pressure. Because PDS-2 containment venting fails, containment failure occurs long-1.05E-1 1

<0.001 term. Late injection is available from low-pressure systems (core spray, LPCI, and fire water) provided they survive containment failure.

Short-term LOCA with loss of high-pressure core makeup and failure to depressurize the primary system for low-pressure PDS-3 core makeup. Core damage occurs at high primary system 8.68E-08 1.35 pressure. Late injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs.

Containment heat removal is available.

Short-term LOCA with loss of high-pressure core makeup, loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core PDS-4 damage occurs at high primary system pressure. Late 0.OOE+00 0.0 injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs. Unlike PDS-3, containment heat removal is unavailable.

Long-term LOCA with loss of high-pressure core makeup and PDS-5 containment heat removal. Core damage occurs at low primary system. Late injection is available from low-pressure 5.59E-101 0.01 systems (core spray, LPCI, and fire water). The containment is vented and intact.

Long-term large LOCA. High-pressure core makeup from HPCI and RCIC are unavailable due to the large LOCA.

Because containment venting fails, containment failure occurs PDS-6 long-term. Late injection is available from low-pressure 0.00E+00 0.0 systems (core spray, LPCI, and fire water) provided they survive containment failure. Core damage occurs at low primary system pressure.

Short-term large LOCA with loss of core cooling. Core PDS-7 damage results at low primary system pressure. Late 1.12E-09 0.02 injection from firewater cross tie and containment heat I

1The revised value replaces the original ER value of 0.0 which was inadvertently submitted.

14 of 68

Table RAI.2-2 Summary of PNPS Core Damage Accident Sequences Plant Damage States removal is available.

Short-term large LOCA with loss of core cooling. Core damage results at low primary system pressure. Late 443E09 0.07 injection from firewater cross tie is available. However, unlike PDS-8 PDS-7, containment heat removal is unavailable.

Short-term LOCA with loss of high and low-pressure core cooling. Because the primary system is depressurized, core damage results at low primary system pressure. Late 3.64E-09 0.06 injection from SSW system, containment venting, and PDS-9 containment heat removal are available.

Short-term LOCA with loss of high and low-pressure core cooling. Because the primary system is depressurized, core damage results at low primary system pressure. Late injection from SSW system and containment heat removal is available. However, unlike PDS-9, containment venting is not PDS-10 available.

Short-term LOCA with loss of high and low-pressure core cooling. Core damage results at low primary system pressure. Late injection from SSW system is available.

0.OOE+00 0.0 However, unlike PDS-9, containment venting and containment PDS-1 1 heat removal are unavailable.

Transient with a loss of long-term decay heat removal. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. The containment is PDS-12 vented and remains intact at the time of core damage.

Transient with a loss of long-term decay heat removal. Core damage results at high primary system pressure. Late in-3.75E-06 58.5 vessel and ex-vessel injection is available. Unlike PDS-12 PDS-13 containment venting fails.

Short-term transient with failure to depressurize the primary system. Core damage results at high primary system 1.52E-07 2.37 pressure. Late in-vessel and ex-vessel injection is available.

PDS-14 Containment heat removal from RHR is available.

Short-term transient with failure to depressurize the primary system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available.

5.07E-08 0.79 Containment heat removal from RHR is available. However, PDS-1 5 containment venting is not available.

Short-term transient with failure to depressurize the primary system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available.

4.89E-09 0.08 Containment heat removal from RHR is not available, but PDS-16 containment venting is available.

Short-term transient with failure to depressurize the primary system. Core damage results at high primary system 2.53E-09 0.04 pressure. Late in-vessel and ex-vessel injection is available.

PDS-17 Neither containment heat removal from RHR nor containment 15 of 68

Table RAI.2-2 Summary of PNPS Core Damage Accident Sequences Plant Damage States venting is available.

Transient with a loss of long-term decay heat removal. Core damage results at low primary system pressure. Late in-1.56E-06 24.40 vessel and ex-vessel injection is available. The containment PDS-1 8 is vented and remains intact at the time of core damage.

Transient with a loss of long-term decay heat removal. Core damage results at low primary system pressure. Late in-vessel and ex-vessel injection is available. Unlike PDS-18 PDS-19 containment venting fails.

Long-term transients with loss of core cooling. Core damage results at low primary system pressure. No late injection, but 6.78E-1 1

<0.001 PDS-20 containment heat removal is available.

Short-term transients (IORV) with loss of core cooling. Core damage results at low primary system pressure. Late 8.18E-09 0.13 PDS-21 injection and containment heat removal are available.

Short-term transients with loss of core cooling. Core damage results at low primary system pressure. Late injection and 1.08E-09 0.02 containment heat removal are available. However, PDS-22 containment venting is not available.

Short-term transients with loss of core cooling. Core damage results at low primary system pressure. Late injection and containment venting are available, but containment heat PDS-23 removal is not available.

Similar to PDS-23, except that containment venting is not 4.98E-09 0.08 PDS-24 available.

4.98E-09 0.08 Short-term transients with loss of core cooling. Core damage results at low primary system pressure. No late injection, but containment heat removal and containment venting are PDS-25 available.

Similar to PDS-25, except that containment venting is not 1.24E-08 0.19 PDS-26 available.

1.24E-08 0.19 Short-term transients with loss of core cooling. Core damage results at low primary system pressure. Late injection and 4.40E-1 1

<0.001 containment heat removal are not available. However, PDS-27 containment venting is available Short-term transients with loss of core cooling. Core damage results at low primary system pressure. Late injection, 1.10E-09 0.02 containment heat removal and containment venting are not PDS-28 available.

Long-term SBO involving loss of injection at high primary system pressure from battery depletion. All accident-mitigating functions are recoverable when AC power is 1.41E-07 2.20 PDS-29 restored.

Short-term SBO sequence involving a loss of high-pressure injection at high primary system pressure from loss of all AC 0.OOE+00 0.00 PDS-30 power and DC power or failure of SRVs. All accident-16 of 68

Table RAI.2-2 Summary of PNPS Core Damage Accident Sequences Plant Damage States mitigating functions are recoverable when offsite power is restored.

Long-term SBO sequence involving a loss of high-pressure injection due to one stuck-open safety relief valve or long-term failure of HPCI/RCIC and subsequent failure to depressurize 2.60E-09 0.04 the primary system. Core damage results at low primary system pressure. All accident-mitigating functions are PDS-31 recoverable when offsite power is restored.

Short-term SBO sequence involving a loss of high-pressure injection due to two stuck-open safety relief valves or failure of HPCI/RCIC and one stuck-open safety relief valve. Core damage results at low primary system pressure. All accident-mitigating functions are recoverable when offsite power is PDS-32 restored.

Short-term large reactor vessel rupture. The resulting loss of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary system 4.00E-09 0.06 pressure. Vessel injection and all forms of containment heat removal (RHR and containment venting) are available. The PDS-33 containment is not bypassed and AC power is available.

Similar to PDS-33, except that containment heat removal from 0.OOE+00 0.00 PDS-34 RHR fails.

Short-term large reactor vessel rupture. The resulting loss of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary system pressure. Vessel injection is unavailable. However, all forms 0.OOE+00 0.00 of containment heat removal (RHR and containment venting) are available. The containment is not bypassed and AC PDS-35 power is available.

Similar to PDS-35, except that containment heat removal from 0.00E+00 0.00 PDS-36 RHR fails.

Short-term ATWS with failure of SRVs/SVs to open to reduce primary system pressure. The ensuing primary system over pressurization leads to a LOCA beyond core cooling 1.95E-08 0.30 capabilities. Late injection and containment heat removal are PDS-37 available.

Short-term AiWS that leads to early core damage at low primary system pressure following successful reactivity 0.00E+00 0.00 control. Late injection is not available. However, containment PDS-38 heat removal is available.

I Similar to PDS-38 except that containment heat removal from 2.32E-09 0.04 PDS-39 the RHR system is not available.

_.3E-9_.0 Long-term ATWS that leads to late core damage at low primary system pressure following successful reactivity 0.00E+00 0.00 control. Late injection is available; containment heat removal PDS-40 from the RHR is not available. The containment is vented.

17 of 68

Table RAI.2-2 Summary of PNPS Core Damage Accident Sequences Plant Damage States Short-term ATWS that leads to early core damage at high primary system pressure following successful reactivity 1.34E-1 1

<0.001 control. Late injection and containment heat removal are PDS-41 available.

Similar to PDS-41 except that containment heat removal from 0.00E+00 0.00 PDS-42 the RHR system is not available.

Long-term ATWS that leads to late core damage at high primary system pressure following successful reactivity 0.00E+00 0.00 control. Late injection is available; containment heat removal PDS-43 from the RHR is not available. The containment is vented.

Long-term ATWS that leads to late core damage at high primary system pressure following successful reactivity control. Late injection is available. However, containment 0.OOE+00 0.00 heat removal from the RHR system and containment venting PDS-44 are not available.

Short-term ATWS that leads to containment failure and early core damage at high primary system pressure because of inadequate reactor water level following a loss of reactivity PDS-45 control. Late injection and containment venting are available.

Short-term ATWS that leads to containment failure and early core damage at high primary system pressure because of inadequate reactor water level following successful reactivity 0.OOE+00 0.00 control. No late injection; however, containment venting is PDS-46 available.

Unisolated LOCA outside containment with early core melt at 322E09 0.05 PDS-47 high RPV pressure.

PDS-48 Unisolated LOCA outside containment with early core melt at 7.73E-10 0.01

__DS-48 low RPV pressure.

7.73E-10_

0.01_

18 of 68

Table RAI.2-3 Grouping of Source Term Magnitude Cesium Iodine (Csl)

Tellurium (Te) Release Fraction Release Fraction 10-4 to 0.001 0.001 to 0.01 0.01 to 0.1 10-4 to 0.01 Low Medium High 0.01 to 0.1 Low Medium High 0.1 to 1.0 Low Medium High 19 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release CAPB Description Bin (CET Category CAPB endstate)

APB-1 Recovered in-vessel, no containment failure NCF CAPB-1

[CD, No VB, No CF, No CCII Reactor pressure vessel (RPV) at low pressure, APB-2 recovered in-vessel, late containment failure, in-vessel Late Low CAPB-12

[CD, VB, Late CE, WW, RPV pressure fission product release goes to torus RPV at low pressure, recovered in-vessel, late APB-3 containment failure, in-vessel fission product release Late Low CAPB-14

[CD, VB, Late CF, DW, RPV pressure mitigated in drywell

<200 psig at VB, No CCI]

RPV at low pressure, recovered in-vessel, late APB-4 containment failure, in-vessel fission product release Late Low CAPB-14

[CD, VB, Late OF, DW, RPV pressure unmitgated<200 psig at VB, No CCI]

unmitigated APB-4A RPV at low pressure, no CCI, early containment failure, Early Low CAPB-9

[CD, VB3, Early CF, DW, RPV APB-4A ex-vessel fission product release not mitigated pressure <200 psig at VB, No CCI]

APB-5 Recovered in-vessel, no containment failure NCF CAPB-2

[CD, VB, No CF, No CCI]

RPV at low pressure, recovered ex-vessel, late APB-6 containment failure, in-vessel fission product release Late Low CAPB-12

[CD, VB, Late OF, WW, RPV pressure goes to torus

<200 psig at VB, No CCI]

RPV at low pressure, recovered ex-vessel, late APB-7 containment failure, in-vessel fission product release Late Low CAPB-14

[CD, VB, Late CE, DW, RPV pressure mitigated in drywell

<200 psig at VB, No CCI]

RPV at low pressure, recovered ex-vessel, late APB-8 containment failure, in-vessel fission product release Late Low CAPB-14

[CD, VB, Late CF, DW, RPV pressure mitigated by the reactor building APB-9 RPV at low pressure, no CCI, late containment failure, Late Low CAPB-14

[CD, VB, Late CE, DW, RPV pressure in-vessel fission product release is unmitigated

__<200 psig at VB, No CCI APB-10 Core Concrete Interactions (CCI) occurs, no NCF CAPB-3

[CD, VB, No CF, CCI) containment failure APB-1 1 RPV at low pressure, CCI occurs, late containment Late Low CAPB-13

[CD, VB, Late CF, WW, RPV pressure failure, in-vessel release goes to torus

<200 psig at VB, CCI]

RPV at low pressure, CCI occurs, late containment

[CD, VB, Late CF, DW, RPV pressure APB-12 failure, in-vessel release mitigated in containment Late Low CAPB-15

<200 psig at VB, CCI]

20 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release CAPB Description Bin (CET Category CAPB endstate)

RPV at low pressure, CCI occurs, late containment APB-13 failure, in-vessel fission product release mitigated by Late Low CAPB-1 5

[CD, VB, Late OF, DW, RPV pressure reactor building

<200 psig at VB, CCl]

APB-14 RPV at low pressure, CCI occurs, late containment failure, in-vessel fission product release not mitigated Late Medium CAPB-15

[CD, VB, Late OF, DW, RPV pressure

___________<200 psig at VB, CCI]

RPV at low pressure, no CCI, early containment failure, APB-15 in-and ex-vessel fission product release mitigated by Early Low CAPB-5

[CD, VB, Early OF, WW, No RPV torus pressure <200 psig at VB, No CCI]

RPV at low pressure, no CCI, early containment failure, APB-1 6 in-and ex-vessel fission product release mitigated by Early Low CAPB-9

[CD, VB, Early OF, OW, No RPV drywell sprays pressure <200 psig at VB, No CCI]

RPV at low pressure, no CCI, early containment failure, APB-17 in-and ex-vessel fission product release mitigated by Early Low CAPB-9

[CD, VB, Early OF, OW, No RPV reactor building pressure <200 psig at VB, No CCl]

RPV at low pressure, no CCI, early containment failure, APB-1 8 in-vessel fission product release to torus, ex-vessel and Early Low CAPB-9

[CD, VB, Early 2

F, OW, No RPV late fission product release not mitigated pressure <200 psig at VB, No CCI]

RPV at low pressure, CCI occurs, early containment APB-1 9 failure, in-and ex-vessel product release mitigated by Early Medium CAPB-7

[CD, VB, Early OF, WW,,RPV torus pressure <200 psig at VB, CCl]

RPV at low pressure, CCI occurs, early containment APB-20 failure, in-and ex-vessel product release mitigated by Early Medium CAPB-1 1 pCD, VB, Early OF, W,,RPV drywell sprays pressure <200 psig at VB, CCl]

RPV at low pressure, CCI occurs, early containment APB-21 failure, in-and ex-vessel product release mitigated by Early Medium CAPB-i1

[CD, VB, Early OF, OW,,RPV reactor building__pressure

<200 psig at VB, CCl]

RPV at low pressure, CCI occurs, early containment APB-22 failure, in-vessel fission product release to torus, ex-Early High CAPB-1 1

[CD, VB, Early CF, W,,RPV vessel and late fission product release not mitigated pressure <200 psig at VB, CCI]

APB-23 RPV at low pressure, no CCI, early containment failure, Early Low CAPB-9

[CD, VB, Early OF, DW, RPV ex-vessel fission product release mitigated by drywell Earlypressure

<200 psig at VB, No CCI]P 21 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release Bin (CET Category CAPB CAPB Description endstate) sprays RPV at low pressure, no CCI, early containment failure, APB-24 ex-vessel fission product release mitigated by reactor Early Low CAPB-9

[CD, VB, Early OF, DW, No RPV building pressure <200 psig at VB, No CCI]

APB-25 RPV at low pressure, no CCI, early containment failure, Early Low CAPB-9

[CD, VB, Early CF, DW, RPV APB-25 ex-vessel fission product release not mitigated pressure <200 psig at VB, No CCI]

RPV at low pressure, CCI occurs, early containment APB-26 failure, ex-vessel product release mitigated by drywell Early Medium CAPB-1 1

[CD, VB, Early OF, W,,RPV sprays pressure <200 psig at VB, CCl]

RPV at low pressure, CCI occurs, early containment APB-27 failure, ex-vessel product release mitigated by reactor Early Medium CAPB-1 1

[CD, VB, Early CF, W,,RPV building pressure <200 psig at VB, CCl]

APB-28 RPV at low pressure, CCI occurs, early containment Early High CAPB-1 1

[CD, VB, Early OF, DW, RPV failure, ex-vessel product release not mitigated

[pressure <200 psig at VB, CCIR APB-29 Recovered in-vessel, no containment failure NCF CAPB-2

[CD, VB, No CF, No CCI]

APB-30 RPV at low pressure, no CCI, late containment failure, Late Low CAPB-12

[CD, VB, Late CF, WW, RPV pressure in-vessel and late fission product release goes to torus

<200 psig at VB, No CCl]

RPV at low pressure, no CCI, late containment failure, APB-31 in-vessel and late fission product release mitigated by Late Low CAPB-14

[CD, VB, Late OF, DW, RPV pressure drywll pras

________<200 psig at VB, No CCl]

drywenl sprays RPV at low pressure, no CCI, late containment failure,

[CD, VB, Late CF, DW, RPV pressure APB-32 in-vessel and late fission product release mitigated by Late Low CAPB-14

<200 psig at VB, No CCR]

reactor building RPV at low pressure, no CCI, late containment failure,

[CD, VB, Late CF, DW, RPV pressure APB-33 in-vessel and late fission product release mitigated by Late Low CAPB-14

<200 psig at VCE, No CCR]

reactor building APB-34 Core Concrete Interactions (CCI) occurs, no NCF CAPB-3

[CD, VB, No CF, CCI) containment failure RPV at low pressure, CC[ occurs, late containment

[CD, VB, Late CE, WW, RPV pressure APB-35 failure, in-vessel and late fission product release Late Low CAPB-13

[CD, VB, at CF, W R

ps I mitigated by torus

<200 psig at VB, CCI]

22 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release Bin (CET Category CAPB endstate)

RPV at low pressure, CCI occurs, late containment

[CD, VB, Late CF, DW, RPV pressure APB-36 failure, in-vessel and late fission product release Late Low CAPB-15

<200 psig at VB, CCID mitigated by drywell sprays RPV at low pressure, CCI occurs, late containment APB-37 failure, in-vessel and late fission product release Late Low CAPB-15

[CD, VB, Late OF, DW, RPV pressure mitigated by reactor building

<200 psig at VB, CCI]

RPV at low pressure, CCI occurs, late containment APB-38 failure, in-vessel and late fission product release not Late Medium CAPB-15

[CD, VB, Late CF, DW, RPV pressure mitigated

<200 psig at VB, CCI]

RPV at low pressure, RPV injection not recovered, no

[CD, VB, Early CF, WW, RPV APB-39 CCI, early containment failure, in-and ex-vessel fission Early Low CAPB-5

[CD, VB, Early CE, DW, RPV product release mitigated by torus pressure <200 psig at VB, No CCI]

RPV at low pressure, RPV injection not recovered, no APB-40 CCI, early containment failure, in-and ex-vessel fission Early Low CAPB-9

[CD, VB, Early CF, DW, No RPV product release mitigated by drywell buidin pressure <200 psig at VB, No CCl]

RPV at low pressure, RPV injection not recovered, no APB-41 CCI, early containment failure, in-and ex-vessel fission Early Low CAPB-9

[CD, VB, Early CF, DW, RPV product release mitigated by reactor building pressure <200 psig at VB, No CCI]

RPV at low pressure, RPV injection not recovered, no APB-42 occurs, early containment failure, in-vandex-velssioneprc Early M m

CAPB-7

[CD, VB, Early CF, DW, RPV releaseto torus,Ct ex-vessel gand late fission fissnproduct pB pressure <200 psig at VB, No CCI]

release not mitigated RPV at low pressure, RPV injection not recovered, CCI APB-43 occurs, early containment failure, in-and ex-vessel Early Medium CAPB-7

[CD, VB, Early CF, DW,,RPV fission product release mitigate by torus pressure <200 psig at VB, CCl]

RPV at low pressure, RPV injection not recovered, CCI APB-44 occurs, early containment failure, in-and ex-vessel Early Medium CAPB-1 1

[CD, VB, Early CF, DW, RPV fission product release mitigate by drywell sprays pesr 20pi tVCI RPV at low pressure, RPV injection not recovered, CCl bidn

V13, Esre <20 p a,

CCI APB-45 occurs, early containment failure, in-and ex-vessel Early Medium CAPB3-11 presur VB, 0

Early CF, V13, RPV fission product release mitigate by reactor building pesr 20pi tVCl 23 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release CAPB Description Bin (CET Category CAPB endstate)

RPV at low pressure, RPV injection not recovered, CCI occurs, early containment failure, in-vessel fission

[CD, VB, Early CF, DW, RPV APB-46 product release to torus, ex-vessel and late fission pressure <200 psig at VB, CCI]

product release not mitigated RPV at low pressure, RPV injection not recovered, no APB-47 CCI, early containment failure, ex-vessel fission product Early Low CAPB-9 pCD, VB, Early CE, DW, RPV release mitigated by drywell sprays pressure <200 psig at VB, No CCI]

RPV at low pressure, RPV injection not recovered, no APB-48 CCW, early containment failure, ex-vessel fission product Early Low CAPB-9

[CD, VB, Early CE, DW, RPV release mitigated by reactor building pressure <200 psig at VB, No CCI]

RPV at low pressure, RPV injection not recovered, no APB-49 CCI, early containment failure, ex-vessel fission product Early Low CAPB-9

[CD, VB, Early CE, DW, No RPV release not mitigated pressure <200 psig at VB, No CCI]

RPV at low pressure, RPV injection not recovered, CCI APB-50 occurs, early containment failure, ex-vessel fission Early Medium CAPB-1 1

[CD, VB, Early CE, W,,RPV product release mitigated by drywell sprays

_pressure

<200 psig at VB, CC]

RPV at low pressure, RPV injection not recovered, CCI APB-51 occurs, early containment failure, ex-vessel fission Early Medium CAPB-1 1

[CD, VB, Early CF, DW, RPV product release mitigated by reactor building pressure <200 psig at VB, CCI]

RPV at low pressure, RPV injection not recovered, CCI APB-52 occurs, early containment failure, ex-vessel fission Early High CAPB-1 1

[CD, VB, Early CE, W,,RPV product release not mitigated pressure <200 psig at VB, CCI]

APB-53 Recovered in-vessel, no containment failure NCF CAPB-2

[CD, VB, No CF, No CCI]

RPV at high pressure, no CCI, late containment failure, APB-54 in-vessel and late fission product release mitigated by Late Low CAPB-12

[CD, VB, Late CF, WW, RPV pressure torus________<200 psig at VB, No CCl]

torus RPV at high pressure, no CCI, late containment failure, APB-55 in-vessel and late fission product release mitigated by Late Low CAPB-14

[CD, VB, Late CF, oW, RPV pressure dryellsprys

_______<200 psig at VB, No CCI]

drywell sprays APB-56 RPV at high pressure, no CCI, late containment failure, Late Low CAPB-14

[CD, VB, Late CF, DW, RPV pressure A in-vessel and late fission product release mitigated by I

1

<200 psig at VB, No CCII 24 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release Bin (CET Category CAPB CAPB Description endstate) reactor building APB-57 RPV at high pressure, no CCI, late containment failure, Late Low CAPB-14

[CD, VB, Late CF, DW, RPV pressure in-vessel and late fission product release not mitigated

<200 psig at VB, No CCI]

APB-58 Core Concrete Interactions (CCI) occurs, no NCF CAPB-3

[CD, VB, No CF, CCI) containment failure RPV at high pressure, CC[ occurs, late containment APB-59 failure, in-vessel and late fission product release Late Low CAPB-13

[CD, VB, Late OF, WW, RPV pressure mitigated by torus

<200 psig at VB, CCII RPV at high pressure, CCI occurs, late containment APB-60 failure, in-vessel and late fission product release Late Low CAPB-15

[CD, VB, Late CF, DW, RPV pressure mitigated by drywell sprays RPV at high pressure, CCI occurs, late containment APB-61 failure, in-vessel and late fission product release Late Low CAPB-15

[CD, VB, Late CE, DW, RPV pressure mitigated by reactor building

<200 psig at VB, CCl]

RPV at high pressure, CCI occurs, late containment APB-62 failure, in-vessel and late fission product release not Late Medium CAPB-15

[CD, VB, Late CF, DW, RPV pressure mitigated

<200 psig at VB, CCl]

RPV at high pressure, RPV injection not recovered, no APB-63 CCI, early containment failure, in-and ex-vessel fission Early Low CAPB-4.

[CD, VB, Early OF, WW, RPV product release mitigated by torus pressure >200 psig at VB, No CCI]

RPV at high pressure, RPV injection not recovered, no APB-64 CCI, early containment failure, in-and ex-vessel fission Early Low CAPB-8

[CD, VB, Early CF, DW, RPV product release mitigated by drywell sprays pressure >200 psig at VB, No CCI]

RPV at high pressure, RPV injection not recovered, no APB-65 CCI, early containment failure, in-and ex-vessel fission Early Low CAPB-8

[CD, VB, Early OF, DW, No RPV product release mitigated by reactor building pressure >200 psig at VB, No CCI]

RPV at high pressure, RPV injection not recovered, no CCI, early containment failure, in-vessel fission product APB-66 release to torus, ex-vessel and late fission product Early Low CAPB-8

[CD, VB, Early CE, DW, RPV release not mitigated pressure >200 psig at VB, No CCI]

25 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release Bin (CET Category CAPB CAPB Description endstate)

RPV at high pressure, RPV injection not recovered, CCI APB-67 occurs, early containment failure, in-and ex-vessel Early Medium CAPB-6

[CD, VB, Early CE, WW, RPV fission product release mitigated by torus pressure >200 psig at VB, CCI]

RPV at high pressure, RPV injection not recovered, CCI APB-68 occurs, early containment failure, in-and ex-vessel Early Medium CAPB-10

[CD, VB, Early OF, OW,,RPV fission product release mitigated by drywell sprays pressure >200 psig at VB, CCI]

RPV at high pressure, RPV injection not recovered, CCI APB-69 occurs, early containment failure, in-and ex-vessel Early Medium CAPB-10

[CD, VB, Early CF, DW,,RPV fission product release mitigated by reactor building pressure >200 psig at VB, CCI]

RPV at high pressure, RPV injection not recovered, CCI occurs, early containment failure, in-vessel fission Early High

[CD, VB, Early CF, DW, RPV APB-70 product release to torus, ex-vessel and late fission CAPB-10pressure >200 psig at VB, CCI]

product release not mitigated RPV at high pressure, RPV injection not recovered, no APB-71 CCI, early containment failure mitigated by drywell Early Low CAPB-8

[CD, VB, Early CF, DW, No RPV sprays pressure >200 psig at VB, No CCI]

RPV at high pressure, RPV injection not recovered, no APB-72 CCI, early containment failure mitigated by reactor Early Low CAPB-8

[CD, VB, Early CE, DW, RPV building pressure >200 psig at VB, No CCI]

RPV at high pressure, RPV injection not recovered, no APB-73 CCI, early containment failure, ex-vessel fission product Early Low CAPB-8

[CD, VB, Early CE, DW, RPV release not mitigated pressure >200 psig at VB, No CCl]

RPV at high pressure, RPV injection not recovered, CCI APB-74 occurs, early containment failure mitigated by drywell Early Medium CAPB-10

[CD, VB, Early CE, DW,,RPV sprays pressure >200 psig at VB, CCl]

RPV at high pressure, RPV injection not recovered, CCI APB-75 occurs, early containment failure mitigated by reactor Early Medium CAPB-10

[CD, VB, Early CE, DW,,RPV building pressure >200 psig at VB, CCI]

RPV at high pressure, RPV injection not recovered, CCI

[CD, VB, Early CF, DW, RPV APB-76 occurs, early containment failure, ex-vessel fission Early High CAPB-1 0 pressure >200 psig at VB, CCl]

product not mitigated p

a I

I 26 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release Bin (CET Category CAPB endstate)

RPV at low pressure, RPV injection not recovered, no BP-D12 CCI, early bypass containment failure, ex-vessel fission Early High CAPB-17 (BYPASS, RPV pressure <200 psig, product release mitigated by drywell sprays No CCII RPV at low pressure, RPV injection not recovered, no BP-D13 CCI, early bypass containment failure, ex-vessel fission Early High CAPB-17 (BYPASS, RPV pressure <200 psig, product release mitigated by reactor building No CCII RPV at low pressure, RPV injection not recovered, no BP-D14 CCI, early bypass containment failure, ex-vessel fission Early High CAPB-17 (BYPASS, RPV pressure <200 psig, product release not mitigated No CCII RPV at high pressure, RPV injection not recovered, no BP-D19 CCI, early bypass containment failure mitigated by Early High CAPB-16

[BYPASS, RPV pressure >200 psig, drywell sprays No CCI]

RPV at high pressure, RPV injection not recovered, no BP-D20 CCI, early bypass containment failure mitigated by Early High CAPB-16 (BYPASS, RPV pressure >200 psig, reactor building No CCl]

RPV at high pressure, RPV injection not recovered, no BP-D21 CCI, early bypass containment failure, ex-vessel fission Early High CAPB-1 6 (BYPASS, RPV pressure >200 psig, product release not mitigated No CCII RPV at low pressure, RPV injection not recovered, CCI BP-E12 occurs, early bypass containment failure, ex-vessel Early High CAPB-19

[BYPASS, RPV pressure <200 psig, fission product release mitigated by drywell sprays CCl]

RPV at low pressure, RPV injection not recovered, CCI BP-E13 occurs, early bypass containment failure, ex-vessel Early High CAPB-19 (BYPASS, RPV pressure <200 psig, fission product release mitigated by reactor building CCl]

RPV at low pressure, RPV injection not recovered, CCI BP-E14 occurs, early bypass containment failure, ex-vessel Early High CAPB-19 (BYPASS, RPV pressure <200 psig, fission product release not mitigated CCl]

RPV at high pressure, RPV injection not recovered, CCI BP-E19 occurs, early bypass containment failure mitigated by Early High CAPB-18

[BYPASS, RPV pressure >200 psig, drywell sprays CCI]

27 of 68

Table RAI.2-4 Description of PNPS CET Release Modes Accident Progression CET Release Mode Description CET Release CAPB Description Bin (CET Category CAPB endstate)

RPV at. high pressure, RPV injection not recovered, CCI BP-E20 occurs, early bypass containment failure mitigated by Early High CAPB-1 8

[BYPASS, RPV pressure >200 psig, reactor building CCI]

RPV at high pressure, RPV injection not recovered, CCI

[BYPASS, RPV pressure >200 psig, BP-E21 occurs, early bypass containment failure, ex-vessel Early High CAPB-1 8 Cals fission product not mitigated 28 of 68

NRC RAI 3 With regard to the treatment and inclusion of external events in the SAMA analysis, provide the following information:

a. The fire CDF (noted as a screening value) has been lowered since the individual plant examination of external events (IPEEE) as a result of updated equipment failure probability and unavailability values. However, the ER states that a more realistic value may be about a factor of three less, or 6.37E-06 per year. Provide a description of the conservatism in the dominant Pilgrim fire CDF sequences (e.g., related to fire initiating event frequencies, severity factors or recovery actions that were not credited) that would support this factor of three.

b. Since the IPEEE, the seismic CDF has been reduced to 3.22x10-5 per year, and is stated to be a conservative value. The ER states that a more realistic value would be a factor of two less, based on engineering judgement. Provide justification to support the factor of two reduction.

c. Entergy's baseline evaluation of SAMA benefits considers only the risk reduction associated with internal events, and neglects the additional risk reduction that a SAMA could have in external events. Entergy does consider the potential for additional risk reduction in external events, but this is done in the context of an upper bound assessment in which the internal event benefits are increased by a factor of six to account for the combined effect of external events and analysis uncertainties. The impact of external events should be reflected in the baseline evaluation, rather than combining the impact of external events with the uncertainty assessment. In this regard, provide a revised baseline evaluation (using a 7 percent discount rate) that accounts for risk reduction in both internal and external events, and an alternate case using a 3 percent discount rate. (Note that the CDF for external events after Entergy's adjustment in the ER is 3.5 times higher than the internal events CDF. This would justify a multiplier of 4.5 or 5, rather than a multiplier of 4 as stated in the ER.)

d. Provide an assessment of the impact on the baseline evaluation results (i.e., the revised baseline evaluation, which accounts for external events) if risk reduction estimates are increased to account for uncertainties in the analysis.

Response to RAI 3a

a. The EPRI Fire Induced Vulnerability Evaluation (FIVE) methodology was used for the original fire IPEEE submittal. This methodology was conservative in several areas, most notably:

" An absence of fire severity factors resulting in conservative estimates of fire frequency.

" Quantification of an older PSA model to obtain conditional core damage probabilities (CCDP) resulting in conservative CDF values.

" No rigorous evaluation of plant operating procedures during fire events resulting in conservative characterization of crew actions.

29 of 68

Simple fire suppression analysis resulting in conservative fire damage and fire spread characterization.

As shown in Table RAI.3-1, accounting for solely the first of these conservatisms by including fire severity factors for dominant fire scenarios 2 and requantifying the CCDP for the transformer fire reduces the fire CDF to 6.11 X 10-6 per year. This value could be further reduced by the addressing the remaining conservatisms listed above. However, these estimate more than justify use of 6.37 x 10.6 per year as the baseline fire CDF.

Response to RAI 3b

b. The updated seismic CDF of 3.22 x 10.5 per reactor-year reflects the updated Gothic computer code room heat up calculations that predict no room cooling requirements for HPCI, RCIC, core spray, and RHR areas; an update of random component failure probabilities; and reduction of the impact of relay chattering in the model to reflect the inherent ruggedness of certain relay type models.

As stated in ER Section 4.21.5.4, conservative assumptions in the updated seismic PSA analysis include the following.

Each of the sequences in the seismic PSA assumes unrecoverable loss of off-site power. If off-site power were maintained, or recovered, following a seismic event, there would be many more systems available to maintain core cooling and containment integrity than are presently credited in the analysis.

" Each of the sequences in the seismic PSA assumes unrecoverable loss of the nitrogen system and the fire water crosstie to the RHR system.

Each of the sequences in the seismic PSA assumes unrecoverable loss of the CSTs water source for the high pressure injection systems.

A single, conservative, surrogate element whose failure leads directly to core damage is used in the seismic risk quantification to model the most seismically rugged components.

Dual initiators are included in the seismic small LOCA, medium LOCA, large LOCA, and ISLOCA event trees. For example, the seismic small LOCA initiating event frequency is a combination of the probability that the seismic event induced a small LOCA and the probability that a small LOCA will occur due to a random event during the 24-hour mission time.

" The ATWS event tree was conservatively simplified so that all conditions which lead to a failure to scram result in core damage, without the benefit of standby liquid control (SLC) or other mitigating systems.

Because there is little industry experience with crew actions following seismic events, human actions were conservatively characterized.

2 Severity factors taken from EPRI, "Fire PRA Implementation Guide", EPRI TR-105928, December 1995, Appendix D.

30 of 68

The updated value is also conservative because it does not credit vessel depressurization via the SRVs following failure to provide high pressure injection from HPCI/RCIC nor include a realistic estimation of the failure to align torus cooling/drywell sprays for containment pressure control. Removing just these two conservatisms as discussed in the sensitivity analysis described below is sufficient to justify reduction of the seismic CDF by a factor of two for the SAMA analysis.

In both the original and the updated seismic PRA, vessel depressurization via the SRVs was not included in the model due to fragility of the nitrogen makeup system to the SRV accumulators. However, the SRVs are available to depressurize the vessel in the short-term. Also, post-seismic operator action to align the backup nitrogen supply to the SRVs is probable given the additional time provided by successful short-term SRV operation.

Hence, this conservatism was removed by modifying the seismic PRA model to include the capability for vessel depressurization via the SRVs.

A realistic estimation of failure to align torus cooling or drywell sprays for containment decay heat removal was derived by using the internal events PSA value of 6.5 x 103 with the following criteria to account for seismic dependencies:

" For seismic hazard levels less than or equal to the seismic design basis earthquake (DBE), the human failure probability was that used in the PSA. It was assumed that a seismic event less severe than the DBE will produce conditions similar to the events addressed in the PSA.

" For seismic hazard level exceeding the DBE (0.15g) but less than 0.5g, the human failure probability is assumed to be twice the PSA value, and ten times the PSA value at 0.5g.

For seismic hazard levels exceeding 0.5g, the human failure probability is predicted to be 0.1 for in-control room human actions and 1.0 for action outside the control room.

Removing these two conservatisms results in a seismic CDF of 1.72 x 10'5 per reactor year, which is a factor of 1.9 reduction in CDF. This value could be further reduced by the addressing the remaining conservatisms. However, this estimate justifies reduction of the seismic CDF by a factor of two for the SAMA analysis.

Response to RAI 3c

c. The SAMA analyses have been redone and presented in the requested format in Table RAI.3-2. As noted in RAI 3c, the appropriate multiplier is 4.51 on the averted cost risk estimates to represent the total SAMA benefits, accounting for both internal and external events. As described in the response to RAI 4c, the core inventory has been revised to account for fuel enrichment and burnup expected during the period of extended operation.

The revised baseline benefit values in Table RAI.3-2 account for both internal and external events conservatively using a multiplier of 5, account for the revised core inventory from response to RAI 4c, and use a 7% discount rate. The 3% discount rate alternate case benefit values in Table RAI.3-2 account for internal and external events using a multiplier of 5, account for the revised core inventory from response to RAI 4c, and use a 3% discount rate.

31 of 68

The revised benefit analyses were performed using Version 1.13.1 of the MELCOR Accident Consequences Code System 2 (MACCS2), which is the latest version of MACCS2 Code, rather than Version 1.12, that was used for the original SAMA analyses provided in the ER.

Version 1.13.1 of the Code corrects errors that had been identified in certain portions of Version 1.12 of the code. The Pilgrim SAMA analysis had not, however, used those portions of Version 1.12 of the Code. Sensitivity analyses have been performed comparing the two versions of the Code which confirm that use of Version 1.13.1 of the Code produces no changes in the Pilgrim SAMA analysis results. Therefore, use of the most recent version of the code is appropriate for the revised analyses.

The estimated costs listed in Table RAI.3-2 reflect the revised values provided in response to RAI # 6b.

Results of the revised baseline analysis show that no additional SAMAs are potentially cost beneficial.

As shown in Table RAI.3-2, no additional SAMAs are potentially cost beneficial with a 3%

discount rate.

Response to RAI 3d

d. The requested information was provided in Table E.2-1. However, the response to RAts 3c and 4c revised the information. As indicated in Section E.1.1 of the ER, CDF uncertainty calculations resulted in a factor of 1.62.

The revised baseline with uncertainty benefit values in Table RAI.3-2 account for both internal and external events using a multiplier of 5, account for revised core inventory from response to RAI 4c, use a 7% discount rate, and account for uncertainty via a 1.62 uncertainty factor. Thus, a factor of 8 is used to account for the combination of the multiplier to account for both internal and external events (5) and the uncertainty factor (1.62).

Results show that no additional SAMAs are potentially cost beneficial even with incorporating an uncertainty factor of 1.62.

32 of 68

Table RAI.3-1 PNPS Dominant Scenarios to Fire CDF Target Ignition Target Damage Unavailability Original Severity New CDF Sarget Frequencya CDF Factor (per year)

Comment F(pereque Probabiliy (CCDP)

(per year)-

(SF) 1 d Scenario 1.90E-02 1.OOE+00 2.90E-05 5.51 E-07 5.51 E-07 HPCI Pump RB6-B 3.40E-04 6.1OE-04 1.00E+00 2.07E-07 0.2 4.15E-08 Oi Fire le Oil Fire le Scenario 2.OOE-02 1.OOE+00 3.70E-05 7.40E-07 0.2 1.48E-07 HPCI Pump Oil Fire TB Heater Bay Area 2b Scenario 4.60E-02 1.00E+00 4.60E-05 2.12E-06 0.2 4.23E-07 Pump Oil Fire Excluding FW Pumps Electrical RA2-A 3.30E-03 1.OOE+00 6.1OE-04 2.01E-06 0.12 2.42E-07 CbetFire 3a Cabinet Fire Scenario 7.50E-03 1.OOE+00 2.40E-06 1.80E-08 1.80E-08 Electrical 4a RA1-F 3.30E-03 1.OOE+00 3.OOE-04 9.90E-07 0.12 1.19E-07 CbetFire Cabinet Fire CR1-A 3.20E-04 1.OOE-01 3.60E-03 1.15E-07 1.15E-07 CR1-B 4.80E-04 1.OOE-01 3.60E-03 1.73E-07 1.73E-07 CR1-C 1.60E-04 1.OOE+00 3.60E-03 5.76E-07 5.76E-07 CR2-A 3.20E-04 1.OOE-01 3.60E-03 1.15E-07 1.15E-07 CR2-B 1.60E-04 1.OOE-01 3.60E-03 5.76E-08 5.76E-08 CR2-C 1.60E-04 1.OOE+00 3.60E-03 5.76E-07 5.76E-07 7

Scenario 5.30E-03 5.OOE-02 3.60E-03 9.54E-07 9.54E-07 9

Scenario 5.50E-03 1.OOE+00 4.40E-04 2.42E-06 0.14 3.39E-07 M-G Set Fire 12 Scenario 7.OOE-03 1.OOE+00 4.40E-04 3.08E-06 0.12 3.70E-07 Electrical Cabinet Fire 33 of 68

Table RAI.3-1 PNPS Dominant Scenarios to Fire CDF 34 of 68

Table RAI.3-2 Revised Summary of Phase II SAMA Analysis Phase Revised Discount II Revised Estimated Baseline Rat SAMA SAMA Baseline Cost Conclusion With Rate SAMA Benefit CostrWity Alternate ID Uncertainty Case Install an independent Not cost 1

method of suppression pool

$234,337

$5,800,000 effective

$374,940

$319,334 cooling.

Install a filtered containment vent to provide fission Not cost 2

product scrubbing. Option

$871,7951

$3,000,000 effective

$1,394,872

$1,218,209 1: Gravel Bed Filter Option 2: Multiple Venturi Scrubber Install a containment vent Not cost 3

large enough to remove

$56,799

>$2,000,000 effective

$90,878

$78,556 ATWS decay heat.

Create a large concrete crucible with heat removal Not cost 4

potential under the basemat $2,405,508 >$100 million effective

$3,848,813 $3,361,353 to contain molten core debris.

Create a water-cooled

$2,405,508 $19,000,000 Not cost

$3,848,813 $3,361,353 rubble bed on the pedestal.

$2,405,508 effective Provide modification for Not cost 6

flooding the drywell head

>$0 0

effective

$0

$0 Enhance fire protection Not cost 7

system and/or SGTS

$59,196

>$2,500,000 effective

$94,714

$82,718 hardware and procedures.

8 Create a core melt source

$2,405,508 >$5,000,000 Not cost

$3,848,813 $3,361,353 reduction system.

effective Install a passive Not cost containment spray system.

$236,327

$5,800,000 effective

$378,123

$321,572 Strengthen primary/

Not cost 10 secondary containment.

$1,151,630 $12,000,000 effective 35 of 68

Table RAI.3-2 Revised Summary of Phase II SAMA Analysis Phase Revised 3%

Revised Estimated Baseline Discount SAMA Baseline Conclusion B

ine Rate SAMA Benefit Cost With Alternate ID Uncertainty Case Increase the depth of the concrete basemat or use an Not cost 11 alternative concrete

$26,907

>$5,000,000 effective

$43,052

$37,599 material to ensure melt-through does not occur 12 Provide a reactor vessel

$5,381

$2,500,000 Not cost exterior cooling system

$2effective

$8,610

$7,520 Construct a building to be 13 connected to primary/

$59,196

>$2,000,000 Not cost

$94,714

$82,718 secondary containment that effective is maintained at a vacuum 14 Dedicated Suppression

$234,337

$5,800,000 Not cost Pool Cooling

$234,337 effective

$374,940

$319,334 15 Create a larger volume in

$1,151,630 $8,000,000 Not cost

$1,842,609 $1,609,238 containment.

effective Increase containment pressure capability Not cost 16 (sufficient pressure to

$1,151,630 $12,000,000 effective

$1,842,609

$1,609,238 withstand severe accidents).

Install improved vacuum Not cost 17 breakers (redundant valves

$0

>$1,000,000 effective

$0

$0 in each line).

18 Increase the temperature

$0

$12,000,000 Not cost

$0

____margin for seals.

effective 19 Install a filtered vent

$871,7951

$3,000,000 Notect

$1,394,872 $1,218,209 effective

$19482$,820 20 Provide a method of drywell

$0

>$1 00000 Not cost

$0

$0 head flooding.

$0_$1_0000 effective 21 Use alternate method of

$59,196

>$2,500,000 Not cost

$94,714

$82,718 reactor building spray.

effective Provide a means of flooding $1,124,723 $2,500,000 effecosti 22 the rubble bed.

NotIct

$1,799,557 $1,571,639 36 of 68

Table RAI.3-2 Revised Summary of Phase II SAMA Analysis Phase Revised 3%

Revised Estimated Baseline Discount SAMA SAMA Baseline Cost Conclusion With Rate SM Benefit CostrWith Alternate ID Uncertainty Case Install a reactor cavity

$2,405,508 $8,750,000 Not cost 23 flooding system.

$24558$,5,0 effective

$3,848,813

$3,361,353 24 Add ribbing to the

$1,151,630 $12,000,000 Not cost containment shell.

effective

$1,842,609

$1,609,238 25 Provide additional DC

$132,726

$500,000 Not cost

$212,362

$183,030 25 battery capacity.

$12,26

$50,0 effective

$212,362

$183,030 26 Use fuel cells instead of

$132,726 >$1 0

Not cost

$212,362

$183,030 lead-acid batteries.

eff ective 27 Modification for Improving

$838,625

$1,953,682 Not cost DC Bus Reliability

_effective 1,341,800

$1,129,635 Provide 16-hour SBO Not cost 28 injection.

$132,726

$500,000 effective

$212,362

$183,030 29 Provide an alternate pump

$248,313

$1 00000 Not cost power source.

effective

$397,301

$342,381 30 AC Bus Cross-Ties

$426,797

$146,120 Potentially

~cost effective$6286 57,0 31 Add a dedicated DC power

$833,243

$3,000,000 Not cost supply.

effective

$1,333,189

$1,122,116 Install additional batteries or

$833,243

$3,000,000 Not cost 32 divisions.

effective

$1,333,189

$1,122,116 33 Install fuel cells.

$132,726 >$1,000,0002 Not cost

$212,362

$183,030 effective 34 DC Cross-Ties

$109,569

$13,000 Potentially

________cost effective $1531 14, 9

35

~~~~~Not cost

$2232 18,0 35 Extended SBO provisions.

$132,726

$500,000 effective

$212,362

$183,030 36 Locate RHR inside

$8,366 Not cost

$13,385

$10,878 containment.

effective Increase frequency of valve Not cost

$40,808

$34,557 leak testing.

$25,505

$100,000 effective 1

38 Improve MSIV design.

$0 n/aS Not cost

$0

$0 effective

$0

$0 37 of 68

Table RAI.3-2 Revised Summary of Phase II SAMA Analysis Phase Revised 3

P1 Revised Reine Discount II Estimated Cocuin Baseline Rt SAMA SAMA Baseline Cost Conclusion With Rate SAMA Benefit CostrWity Alternate ID Uncertainty Case Install an independent Not cost 39 diesel for the CST makeup

$0

$135,000 effective

$0

$0 pumps.

Provide an additional high Not cost 40 pressure injection pump

$102,606 >$1,000,00 effective

$164,170

$137,423 with independent diesel.

.41 Install independent AC high

$102,606

$1,000,00 Not cost

$164,170

$137,423 pressure injection system.

effective 42 Install a passive high

$102,606 >$1 N cost

$164,170

$137,423 pressure system.

$0,0>10,0_N

_cost 43 Improved high pressure

$68,736

>$1 o

Not cost

$109,977

$91,989 systems effective Install an additional active 2

Not cost

$164,170

$137,423 44 high pressure system.

$102,606 >$1,000,00 e fective Add a diverse injection

$102,606 >$1,000,00 Not cost system.

effective

$164,170

$137,423 46 Increasebilit.

$47,618

$1,800,00re Not cost

$76,188

$63,832 reliability.

effective

$7,18_6383 47 Install an ATWS sized vent.

$56,799

>$2,000,000 Not cost

$90,878

$78,556 eff ective

$9,8

$756 Diversify explosive valve

$0

>$200,000 Not cost 48 operation.

effective

$0 Increase the reliability of Not cost 49 SRVs by adding signals to

$31,881

>$1,500,000 effective

$51,010

$43,196 open them automatically.

50 Improve SRV design.

$172,744

$1,500,000 Not cost effective

$276,391

$232,454 51 Provide self-cooled ECCS

$29,891

>$200,000 Not cost

$47,826

$40,957 pump seals.

effective 2

Provide digital large break

>$1 00,000 Not cost 52 LOCA protection.

$995

>$100,000 effective

$1,592

$1,119 38 of 68

Table RAI.3-2 Revised Summary of Phase II SAMA Analysis Phase Revised 3%

Revised Estimated Baseline Discount SAMA Baseline Conclusion Rate SAMA Benefit Cost With Alternate ID Uncertainty Case Control containment venting Not cost 53 within a narrow band of

$114,364

$300,000 effective

$182,982

$153,582 pressure Install a bypass switch to bypass the low reactor Not cost 54 pressure interlocks of LPCI

$23,515

$1,000,000 effective

$37,624

$32,318 or core spray injection valves.

Improve SSW System and Not cost

$535353

$459,971 RBCCW pump recovery.

$334,596

>$5,000,000 efective Provide redundant DC Potentially 56 power supplies to DTV

$200,010

$112,400 oteti

$320,016

$264,600 valves.

~~~cost effective$3016 24,0 valves.

Proceduralize the use of diesel fire pump 57 hydroturbine in the event of

$156,828

$26,000 Potentially EDGA ailreorcost effective

$250,925

$214,544 EDG A failure or unavailability.

Proceduralize the operator 58 action to feed B1 loads via Potentially B3 when A5 is unavailable

$175,142

$50,000cost effective

$280,226

$236,616 post-trip.

Provide redundant path from fire protection pump

$845,784

$1,956,000 Not cost discharge to LPCI loops A effective

$1,353,255

$1,166,976 and B cross-tie.

39 of 68

Note to Table RAI.3-2:

1. The baseline benefit value of zero dollars submitted in the ER has been revised. The revised value corrects the inadvertent use of the baseline source terms rather than reduced source terms in the original benefit estimate for SAMAs 002 and 019.

In the revised analysis, the benefit of additional filtering capability is estimated by reducing the source terms (by a factor of two) for core damage sequences associated with a loss of containment heat removal (TW sequences), successful containment vent, and loss of reactor vessel makeup occurring some time following vent initiation. Specifically, the source terms associated with core damage accident sequences that are binned into plant damages states 1, 5, 12, 18, 40 and 43 (Table RAI.2-2) are reduced by a factor of 2. These plant damages states are considered 'TW' sequences and the impact on source terms are manifested in late release accident progressions. These are CAPBs 12, 13, 14 and 15. The source terms for the other accident sequences would remain the same.

A comparison between the 'base case' source terms release fractions and the revised source terms release fractions for the filtered containment vent case (SAMA-2) for noble gases, iodine, cesium and tellurium, four of the nine source terms used as input for the MASSC2 code, are presented below.

NG I

Cs Te CAPB-1 2Base 2.0E-01 5.8E-05 4.4E-05 1.3E-07 CAPB-12SAMA-2 2.OE-01 5.7E-05 4.2E-05 1.2E-07

%Change 0.0%

-3.1%

-4.1%

-6.4%

CAPB-1 3 Bas 1.OE+00 8.0E-03 6.0E-03 1.8E-04 CAPB-1 3 SAMA-2 1.OE+00 8.OE-03 6.OE-03 1.8E-04

%Change 0.0%

0.0%

CAPB-14Base 7.8E-01 2.9E-02 2.7E-02 2.5E-05 CAPB-1 4 SAMA-2 7.8E-01 2.9E-02 2.7E-02 2.5E-05

%Change 0.0%

0.0%

CAPB-1 5 Base 1.OE+00 2.8E-01 2.7E-01 1.3E-03 CAPB-1 5 SAMA-2 1.OE+00 1.5E-01 1.4E-01 1.OE-04

%Change 0.0%

-46.7%

-46.6%

-91.8%

Running the MACCS2 code with the reduced source terms results in an 18.5 percent reduction in off-site population dose risk attributable to implementing this SAMA. The use of ER values for averted risk results in a revised baseline benefit of approximately $871,795.

Given that the estimated cost for implementing this SAMA is $3 million, this revised value does not alter the ER conclusion that SAMAs 002 and 019 are 'Not cost effective'.

In redoing the SAMA analyses as requested by this RAI, it was confirmed that SAMA 2 (and its companion SAMA 19) are the only SAMAs that result in reduced source terms such that this correction causes no changes in the other SAMA analyses.

2. The estimated cost reflects the revised value provided in response to RAI #6b.

3. The estimated cost reflects the value provided in response to RAI #5e.

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NRC RAI 4 Provide the following information concerning the MACCS analyses:

a. Annual meteorology data from the year 2001 were used in the MACCS2 analyses. Provide a brief statement regarding the acceptability of use of this year's data rather than a different year's data.

b. For the emergency response assumptions, indicate what percentage of the population was assumed to evacuate.

c. The MACCS2 analysis for Pilgrim is based on a core inventory from a mid-1 980 analysis, scaled by the power level for Pilgrim. Current boiling water reactor (BWR) fuel management practices use longer fuel cycles (time between refueling) and result in significantly higher fuel burnups. The use of the older BWR core inventory instead of a plant specific cycle could significantly underestimate the inventory of long-lived radionuclides important to population dose (such as Sr-90, Cs-134 and Cs-137), and thus impact the SAMA evaluation. Justify the adequacy of the SAMA cost benefit evaluation given the fuel enrichment and burnup expected at Pilgrim during the renewal period.

Response to RAI 4a

a. The 2001 meteorological data set was the most current and complete at the time of data collection for this study. The data were derived from measurements at the two meteorological towers on site.

Response to RAI 4b

b. For the emergency response assumptions, the entire population (or 100% of the population) within the 10-mile emergency planning zone was assumed to evacuate.

Response to RAI 4c

c. Best-estimate inventory of long-lived radionuclides such as Sr-90, Cs-1 34, and Cs-137 were derived from an ORIGEN calculation assuming 4.65% enrichment and average burnup according to the expected fuel management practice. It was found that the best-estimate inventory differed from the power-scaled reference inventory by approximately 25%.

The revised baseline benefits reported in response to RAI 3.c include the impact of the 25%

increase in the inventory values for Sr-90, Cs-1 34, and Cs-1 37 for each analysis case.

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NRC RAI 5 Provide the following with regard to the SAMA identification and screening processes:

a. Table 3-15 of the IPEEE submittal provides a listing of important seismic faults. While no importance values are provided, a number of these faults appear to involve equipment for which some strengthening may be relatively inexpensive. Also, as indicated in the IPEEE and the staff safety evaluation report on the IPEEE, the diesel generator building was found to have limiting fragilities that could significantly impact the CDF. Discuss and evaluate, as necessary, the potential for cost-beneficial SAMAs based on this listing and the known diesel generator building weaknesses.

b. The IPEEE submittal (page 3-44) states that the seismic PRA assumed that low ruggedness relays judged essential under A-46 had been replaced. ER Section E.1.3.1 indicates that the recent reevaluation of seismic risk included the replacement of certain relays with seismically rugged models. Explain this apparent contradiction.

c. ER Table E.1-12 includes a list of the contributors to the updated fire CDF. A number of these have CDF values significantly above 1E-06 per year. For each fire area or dominant fire sequence, explain what measures were taken to further reduce risk, and explain why the fire CDFs cannot be further reduced in a cost effective manner.

d. ER Section E.2.1 states that several enhancements from the IPE or IPEEE were recommended and implemented and that these were included as Phase I SAMA candidates 248 through 281. Provide a detailed accounting of the potential enhancements from the IPE and IPEEE. For each enhancement, indicate if the improvement has been implemented, is no longer being considered and why, and if credit is taken for the improvement in the current PSA. For those enhancements not implemented, indicate their importance and why they should not be considered as Phase II SAMA candidates.

e. Loss of direct current (dc) bus initiators contribute almost 50 percent of the CDF. The only SAMA that directly addresses improving existing dc system reliability is Phase II SAMA 27 and this SAMA reduced CDF by less than 5 percent. Discuss the loss of dc initiators in more detail, their major causes, and the potential for other modifications to reduce the CDF.

f. ER Table E.1-3 indicates that Phase I SAMAs, including procedure and instrumentation improvements, have been implemented to address event FXT XHE-FO-V4T2 (and FXT-XHE-FO-DWS). In spite of these improvements, this event is the highest risk reduction worth ranked non-initiator event. The Phase II SAMAs (57 and 59) cited do not appear to effectively address this event which is an operator error. Identify and evaluate other SAMAs that might lower the importance of this event.

g. ER Table E.1-3 indicates that Phase II SAMA 45 was considered to address event FXT-ENG-FR-P140. This SAMA includes the addition of an entire new system. The addition of a redundant diesel fire pump would appear to be more cost effective. Provide an evaluation of the costs and benefits of adding a redundant diesel fire pump, in lieu of Phase II SAMA 45.

h. ER Table E.1-3 indicates that Phase II SAMA 53 was evaluated to address event CIV XHE-FO-DTV (operator fails to vent containment). This SAMA, controlling containment venting within a narrow pressure band, would be subject to the same failure to vent human error as 42 of 68

in the basic event. Conversion of the containment vent system to a passive design would appear to be more effective in reducing the risk from this event. Provide an evaluation of the costs and benefits of converting the vent system to a passive design.

Response to RAI 5a

a. IPEEE Table 3-15 lists the important basic events (seismic faults and operator actions) that dominate seismic risk. Many of the components listed are important due to their physical correlation in relation to redundant equipment, e.g., correlated seismic failure of all RBCCW pumps, all RHR pumps, all SSW pumps, MCC B14 & MCC B15, etc. Relocation of such equipment would be cost prohibitive. Several other components have high seismic capacity as indicated in IPEEE Table 3-11 (median capacities of 1.0g PGA or greater) and no measurable benefit would be expected from further strengthening. The only component (other than piping) listed in IPEEE Table 3-11 with a median capacity < 1.0 PGA is the EDG building.

The purpose of the IPEEE was to identify plant vulnerabilities and if necessary, reduce the overall likelihood of core damage and radioactive material release by modifying hardware and procedures to help prevent or mitigate severe accidents. As discussed in the IPEEE report, the north wall of the EDG building concrete structures emerged as having a potentially limiting fragility for above design basis accidents. This prompted a search for an alternative source of emergency power. After the initial sensitivity testing of the seismic PRA, the Station Blackout (SBO) diesel was introduced into the model as an alternative source of emergency power. Walkdowns of the SBO diesel revealed a potential weakness in the support of the muffler. The longitudinal direction of the noise suppression muffler supports were stiffened and subsequent analysis of the SBO diesel resulted in acceptable ruggedness for this component. Since the SBO diesel provides an alternative source of emergency power for above design basis accidents, no further evaluation of the identified EDG building fragility is necessary.

In addition, the IPEEE seismic study was conservative because it did not credit certain systems capable of mitigating severe accidents. For example, the LPCI mode of the RHR system was not credited, nor was low pressure injection from the fire water pump cross-tie.

Seismic core damage risk could be lowered significantly by inclusion of these systems.

Response to RAI 5b

b. Low ruggedness relays judged essential under A-46 were assumed to be replaced, hence, the IPEEE seismic analysis was done with rugged relays assumed in the model. In addition, the seismic PRA continued to evaluate the potential impact of relay chattering on systems performance during a seismic event. The seismic PRA looked at the response of systems, structures and components over a wide spectrum of seismic events, and some rugged relays would be expected to chatter at high enough g forces. These chattering relays could fail some functions, and these failures were included as basic events.

The relay chatter analysis looked for events which could prevent certain seismic PRA functions from occurring. For each of these relays, a basic event for relay failure due to relay chattering was included in the seismic PRA model. These relays were uniquely identified by basic event code, a median capacity, logarithmic standard deviation and ground motion level in which the relay is assumed not to fail.

43 of 68

During the re-evaluation of the seismic PRA model, since all essential relays were judged to be of high ruggedness, the impact of relay chattering was reduced by increasing the median seismic capacity to 2.0g to reflect the inherent ruggedness for these relay type models.

Therefore, the statement, "and to model replacement of certain relays with a seismically rugged model," in ER Section E1.3.1 is hereby revised to state, "and to reduce the impact of relay chattering to reflect the inherent ruggedness for certain relay type models."

Response to RAI 5c

c. There were five zone scenarios which produced fire CDF contributions in excess of 1.0 X 10-6 per year. These were due to modeling conservatism and were each addressed and reduced in Table RAI.3-1 presented in the response to RAI3a. Since the applied severity factors reduced individual contributions to below the 1.0 X 10.6 per year threshold, modifications to further reduce the CDF would not be cost effective.

The risk significant fire areas are equipped with a detection system that alarms in the control room. Also, several zones are equipped with a suppression system. Therefore, no cost-effective hardware changes were identified to reduce CDF in these areas.

In addition, the Fire Protection Program uses a three-tiered approach:

1. Prevent fires from starting.

2. Detect fires promptly, suppressing them quickly, and thereby limiting fire damage.

3. Design plant safety systems so that a fire which does start will not ultimately prevent essential plant safety functions from being accomplished.

Following the Fire Hazards Analysis provisions and Fire Protection Program procedures provides assurance that risk in these areas is minimized. Therefore, no cost-effective procedural changes were identified to reduce CDF in these areas.

Response to RAI 5d

d. Table RAI.5-1 presents Phase I SAMAs candidates 248 through 281, which are the enhancements recommended in the IPE, PSA update, and IPEEE. Those with a reference source labeled [17] are from the IPE or PSA update. Those with a reference source labeled

[18] are from the IPEEE.

Phase I SAMAs candidates 248 through 260 are the key human actions identified to enhance safety and reduce risk. These human actions have been implemented. In addition, an operator lesson plan is used for continuing training of operators on PSA insights and important operator actions.

Phase I SAMA candidates 262, 267, and 268 have also been implemented and included in the current 2003 PSA model. However, SAMA 269 is not considered in the 2003 PSA model update. As indicated in the response to RAI 1 b, SAMA 269 modification will slightly reduce CDF, but would not impact the results of the SAMA analysis.

Phase I SAMA candidates 272 through 281, from the IPEEE, have been implemented.

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Phase I SAMAs 261, 263, 264, 265, 266, 270, 271 were retained as Phase II SAMA candidates.

Response to RAI 5e

e. The most important contributor to CDF involves loss of one or both divisions of 125V DC power. Although loss of one DC bus is not likely to result in an automatic scram, procedures direct the plant to be manually shut down if immediate repairs to the bus are not imminent.

Loss of a single DC bus results in loss of various DC distribution panels and loss of control power to a 4160-VAC safeguard bus (DC power for control and/or operation of a diesel generator, RCIC or HPCI, one loop of torus cooling and drywell sprays, one train of LPCI and core spray injection). The dominant loss of DC sequences are accompanied by loss of the suppression pool cooling and drywell spray modes of RHR and subsequent loss of containment heat removal.

Because PNPS operating history has no occurrences of loss of a DC bus, loss of a safety DC bus initiator was calculated by quantifying a simple reliability model. A fault tree was constructed to accurately represent the physical plant configuration. The data in the fault tree, i.e. bus failures, reflects current updated plant data. However, the quantification process used only a single value to represent the loss of a DC bus (2.63E-3/ry).

Evaluation of SAMA 27 involved potential modifications to improve DC bus reliability. As a result, changes to DC bus faults failure probabilities were made in the PSA model, which resulted in a 4.65 percent reduction in CDF. In response to this RAI, SAMA 27 was re-evaluated by eliminating the occurrence of a loss of a 125-Vdc bus B initiator. This resulted in a 24.3 percent reduction in CDF and a revised baseline benefit of approximately

$838,625. The cost of installing a new DC source capable of powering both 125VDC busses is estimated to be $1,953,682 by engineering judgment. Therefore, this is not cost effective for PNPS.

However, SAMA 34 (enhancement of plant procedures to cross-tie DC buses) improves DC reliability and was found to be potentially cost beneficial. Also, in response to RAI 7a, a SAMA to power the battery chargers via the security diesel generator was found to be potentially cost beneficial. Therefore, no additional modifications were examined to reduce the loss of a DC bus initiator.

Response to RAI 5f

f.

Events FXT-XHE-FO-V4T2 and FXT-XHE-FO-DWS represent operator failure to align fire water via the LPCI injection path for alternate reactor vessel injection and for alternate drywell spray. To mitigate these failures, a new SAMA is proposed to change the removable spool piece to permanent piping and provide the capability to open locked-closed manual valves 10-HO-511 and 8-1-56 (see Figure RAI.5-1) remotely from the control room.

These modifications would increase the success probability of the actions to align fire water to the LPCI injection path. To assess the benefit of this SAMA, the human error probability (HEP) for FXT-XHE-FO-V4T2 was reduced from 2.31 E-2 to 5.OE-3 and the HEP for FXT-XHE-FO-DWS was reduced from 2.21 E-2 to 5.1 E-3. In addition, this SAMA also changed the manual valves 1 0-HO-51 land 8-1-56 to motor operated valves. Therefore, the probability of manual valve failing to open (5.OE-4) was changed to the probability of motor operated valve failing to open (3.OE-3) for these events. These changes resulted in a CDF 45 of 68

reduction of 2.60% and a revised baseline with uncertainty benefit of approximately

$313,442. The cost of implementing this SAMA is estimated to be $2,860,445 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

Response to RAI 5q

g. A cost benefit analysis was performed to evaluate the addition of a redundant diesel fire pump to address event FXT-ENG-FR-P140, Diesel Fire Pump P140 Fails to Run. A bounding analysis was performed by setting the events for failure of the diesel fire pump to start and to run to zero in the PSA model, which resulted in the CDF reduction of 3.91% and a revised baseline with uncertainty benefit of approximately $654,306. The cost of implementing the addition of a redundant diesel fire pump is estimated to be $5,507,336 by engineering judgment. Therefore, the addition of a redundant diesel fire pump is not cost effective for PNPS.

Response to RAI 5h

h. Phase II SAMA 53 proposes a method to control containment venting within a narrow band to preclude net positive suction head concerns for pumps taking suction from the torus. The evaluation for SAMA 53 considered a reduction in the probability of failure to perform direct torus venting when required. Since SAMA 53 does not directly address plant operators failing to vent containment, a new SAMA is proposed to evaluate the cost benefit of a passive design direct torus vent instead of the existing direct torus vent.

The existing direct torus vent line originates downstream of torus purge exhaust isolation valve AO-5042B and aligns flow to the normally-closed direct torus vent isolation valve AO-5025 (see Figure RAt.2-2). Downstream of AO-5025, the vent line continues to rupture diaphragm PSD-8180. Torus air space pressure must be greater than or equal to 30 psig in order to break the rupture disc associated with this vent path. After rupturing PSD-8180, flow from the direct torus vent line passes to the common outlet from both standby gas treatment trains and proceeds to the main stack. Air operator valves AO-5042B and AO-5025 and associated solenoid valves are controlled from the control room. These valves need 125-VDC power and an air or nitrogen supply to open. They fail closed on loss of air and nitrogen or on loss of power.

The proposed SAMA would modify the air operated valves and the associated solenoid valves so that the air operated valves fail open on loss of air and nitrogen or on loss of power.

Conversion of the existing direct torus vent to a passive torus vent resulted in a CDF reduction of 14.5 percent benefit and a revised baseline with uncertainty benefit of approximately $1,152,242. The cost of changing the direct torus vent to a passive design is estimated to be $3,161,837. Therefore, this SAMA is not cost effective for PNPS.

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Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference Improvements Related to IPE, IPE Update& IPEEE Insights 248 Improve operator

[17]

This SAMA would provide

3 - Already This operator action is taken in response to PNPS procedure action: Align fire addition availability of water installed a loss of high-pressure injection (feedwater, 5.3.26, RPV water cross tie for for RPV injection HPCI, and RCIC) and preferred low Injection During injection via LPCI pressure injection (condensate, LPCI, and Emergencies core spray). It entails installing a spool piece/strainer, opening cross tie valves 10-HO-511 and 8-1-56, starting firewater pump, and opening the LPCI injection valve. This Ihas already been implemented at PNPS.

249 Operator Action:

[17]

This SAMA would provide

3 - Already This operator action is taken in the event PNPS procedure Vent Containment containment pressure control installed that containment heat removal via 5.4.6, Primary Using Direct and containment heat suppression pool cooling and drywell spray Containment Tows Vent removal capability is unavailable. The action entails defeating Venting and the isolation signal for AO-5042B, installing Purging Under fuses for direct tows vent valve AO-5025, Emergency and opening AO-5042B and AO-5025. This Conditions has already been implemented at PNPS.

250 Operator Action:

[171 This SAMA would provide

3 - Already This operator action is taken in the event PNPS procedure Align Fire Water containment pressure control installed that containment heat removal via the RHR 2.2.19.5, RHR Cross tie for and containment heat System (suppression pool cooling and Modes of Operation Drywell Spray removal capability drywell spray) is unavailable. The action for Transients entails aligning the fire water cross tie to LPCI and opening the drywell spray valve.

This has already been implemented at PNPS.

251 Operator action:

[17]

This SAMA would provide

3 - Already This operator action is taken in response to PNPS procedure Align drywell containment pressure control installed events involving loss of the Power 2.2.19.5, RHR spray mode of and containment heat Conversion System (PCS) and Modes of Operation RHR removal capability unavailability of suppression pool cooling, for Transients The action primarily entails opening a RHR heat exchanger bypass valve, and opening the drywell spray valves. This has already been implemented at PNPS.

47 of 68

Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference 252 Operator Action:

[17]

This SAMA would provide

3 -Already This operator action is taken in response to EOP-3, "Primary Initiate containment pressure control installed provide containment heat removal during Containment Suppression pool and containment heat transients, LOCAs and ATWS. The action Control" cooling removal capability entails to align the RHR System suppression pool cooling path for containment heat removal. This operator action has already been implemented at PNPS EOP-3, "Primary Containment Control.

253 Operator Action:

[17]

This SAMA would prevent

3 -Already This operator action is taken in response to EOP-1, "RPV Manually initiate the core damage during installed depressurize the reactor to allow the low Control' emergency transients, small and medium pressure injection systems to provide depressurization LOCAs, and ATWS coolant makeup to the reactor pressure vessel during transients, small and medium LOCAs, and ATWS. This operator action has already been implemented at PNPS EOP -1, "RPV Control".

254 Operator action:

[17]

Availability of additional

3 -Already This operator action is taken in response to PNPS Procedure Align Station electric power sources for installed align the Station Blackout diesel generator 5.3.31 Station Blackout diesel coping with the loss of if a loss of offsite power were to occur. The Blackout generator normal power Station Blackout diesel generator can be aligned to provide electric power to 416OVac busses A5 or A6. With bus B1 or B2 energized and supplying MCC B15 or B14 and B20 battery charging is maintained and RHR valves necessary for aligning the diesel fire pump for RPV vessel injection can be remotely operated. This operator action has already been implemented at PNPS Procedure 5.3.31, Station blackout 2.2.146, 'Station Blackout Diesel Generator", and 2.416 "Distribution Alignment Electrical System Malfunction" 255 Operator Action:

[17]

This SAMA would reduce the #3 - Already This operator action is taken to restore Procedure 2.4.16, Recovery of core damage frequency installed orfsite power within 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months from the start Distribution offsite power contribution from the loss of a loss of offsite power event and Alignment Electrical within 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months subsequent loss of onsite AC power System 48 of 68

Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference during loss of offsite power event resulting in a plant station blackout.

Malfunctions normal power Procedure 2.4.16, attachmentl 1, event Restoration of AC Power provides guidance in restoring offsite AC power to the plant.

Additionally, attachment 12 provides a flow chart for optimizing AC Power Restoration.

256 Operator Action:

[17]

This SAMA would provide

3 -Already This operator action is taken in response to EOP-2, RPV Initiate Standby boron injection during ATWS installed provide boron injection ATWS. The Control Failure-to-liquid control operator action entails to align the standby Scram system liquid control system for boron injection.

This operator action has already been implemented at PNPS EOP-2, RPV Control Failure-to-Scram.

257 Operator Action:

[17]

Availability of alternate

3 -Already The fire water system can be crosstied to 5.3.26, RPV Align alternate external water injection to the installed feedwater system for long-term recovery Injection During RPV injection RPV through the use of the RPV injection. This requires the use of a Emergencies using Condensate Condensate and Feed water Plymouth fire pump truck on site. This and Feedwater System post core damage operator action has already been System with implemented at PNPS.

service water makeup 258 Operator Action:

(17]

This SAMA would impact

3 -Already This operator action is taken in response to SAG-01" RPV and Manually initiate accident sequence timing, installed depressurize the reactor to allow restoration Primary RPV and the occurrence of severe of low-pressure injection to a damaged Containment depressurization accident phenomena that core, the potential elimination of severe Flooding" post core damage challenges containment accident phenomena related to direct integrity containment heating (DCH) and the potential reduction in radionuclide releases.

For example, if the RPV is breached at high pressure, venting into an open containment, the radionuclide releases are substantially higher than with the RPV depressurized at the time of failure. This operator action has already been implemented at PNPS in SAG-01" RPV and Primary Containment Flooding".

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Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference 259 Operator action:

[17]

This SAMA would provide

3 - Already This operator action is taken flood the RPV PNPS procedure Align firewater addition availability of water installed and primary containment during a severe 5.3.26, RPV cross tie for for RPV injection post core accident. Plant operators initiate firewater Injection During injection via LPCI damage as directed by SAG-system injection to the RPV prior to vessel Emergencies post core 01" RPV and Primary failure given that there is water in the fire SAG-01" RPV and damage.

Containment Flooding" water storage tanks and that power is Primary available. The action entails installing a Containment spool piece/strainer, opening crosstie Flooding valves 1 0-HO-51 1 and 8-1-56, starting a firewater pump, and opening a LPCI injection valve.

This operator action has already been implemented at PNPS in SAG-01" RPV and Primary Containment Flooding".

260 Operator action:

[17]

This SAMA would provide

3 - Already This operator action is taken in response to PNPS procedure Align drywell containment pressure installed events that occur post core damage that 5.3.26, RPV sprays post core control, containment radiation result in high drywell pressure conditions, Injection During damage control and primary high radiation in either the drywell or tours, Emergencies containment flooding or the requirement for containment flooding. PNPS procedure capability The action primarily entails starting an RHR 2.2.19.5, RHR pump or fire water system pump, closing Modes of Operation the RHR heat exchanger bypass valve, and for Transients opening the drywell sprays valves. This SAG-01" RPV and operator action has already been Primary implemented at PNPS in SAG-01" RPV and Containment Primary Containment Flooding".

Flooding 261 Control

[17]

This SAMA would establish a Retain Procedural changes and training will be containment

[5]

narrow pressure control band required to implement this SAMA.

venting within a to prevent rapid containment narrow band of depressurization when pressure venting is implemented thus avoiding adverse impact on the low pressure ECCS injection systems taking suction from the tows.

262 Install nitrogen

[17]

This SAMA would improve

3 - Already The containment vent function is the last SDBD-09A, 50 of 68

Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference bottles as backup the nitrogen supply reliability installed resort methods currently specified in BWRs Primary air supply for and extend direct torus vent to remove heat from containment and Containment direct torus vent valve operation time.

control containment pressure under Atmosphere valves extremely adverse cimumstances. Pilgrim Control System has redundant and diverse nitrogen supplies to back up the instrument air supply to the direct torus valves.

263 Install a bypass

[17]

This SAMA would reduce the Considering a modification to Install a switch to bypass core damage frequency Retain bypass switch to bypass the low reactor the low reactor contribution from the pressure interlocks of LPCI or core spray pressure transients with stuck open injection valves. This modification would interlocks of LPCI SRVs or LOCAs cases. Core permit these valves to open for injection.

or core spray Spray and LPCI injection injection valves valves require a low permissive signal from the same two sensors to open the valves for RPV injection.

264 Increase the

[17]

This SAMA would reduce Retain The SSW, and RBCCW systems have reliability of salt common cause redundant loops, each having a minimum of service water dependencies from SWS and two pumps. Adding additional trains of (SSW) and RBCCW Systems and thus equipment to improve pump recovery would RBCCW pumps reduce plant risk through be expensive, costing far more than the system reliability associated risk benefit. Retain for improvement, evaluation.

265 Provide

[17]

This SAMA would improve Consider powering the two series valves, 5.4.6, Primary redundant DC the reliability of the direct Retain AO-5025 and AO-5042B, from DC A or DC Containment power supplies to torus vent valves and B, whichever is live. This would require Venting and direct torus vent enhance the containment adding four fuses to C7, so the operator still Purging Under valves heat removal capability has to inset two (or perhaps four) fuses to Emergency enable the direct torus vent function, as he Conditions does now. Indication of the live bus(s) might be added as well if necessary.

Retain for evaluation.

266 Proceduralize the

[17]

This SAMA would increase The hydroturbine fuel oil transfer pump (P-PNPS procedures:

use of diesel fire applicability of diesel fire Retain 181) has the capability to provide makeup 2.2.25, Fire Water pump pump hydroturbine-driven to the fire pump day tank to allow continued Supply System hydroturbine in fuel pump to reduce the core operation of the diesel fire pump, without 51 of 68

Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through.281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference the event of EDG damage frequency.

dependence on electrical power. However, 2.4.54, Loss of All A failure or existing procedures only direct use of the Fire Suppression unavailability hydroturbine FOTP in the event of a station Pumps, or Loss of blackout (SBO). The primary reason for Redundancy in the this is that operation of the hydroturbine Fire Water Supply FOTP requires isolation (and therefore System unavailability) of EDG A FOTP P-1 41 A.

Thus, for non-SBO sequences involving a loss of offsite power (LOSP) and failure of either EDG A or EDG A FOTP P-141, use of the hydroturbine is not credited.

Procedures could be revised to allow use of the hydroturbine in the event that EDG A or EDG A FOTP P-1 41A is unavailable to reduce the core damage frequency.

267 Proceduralize the

[17]

This SAMA would provide the Currently Pilgrim Breaker Interchangeability Procedure 2.2.7 operator action to direction to close the 480V

3 -Already Matrix provides location of the nearest 480Vac system, Manually Close Circuit Breaker to support installed interchangeable breaker to the one that is E43, sh. 2, Breaker 480V Circuit associated loads failing to remain closed, whether it is in an Interchangeability Breaker adjacent load center or the warehouse.

Matrix, 480V The existence and application of this matrix System will be made known to the operators via appropriate training.

268 Proceduralize the

[17]

This SAMA would provide the Currently Pilgrim Breaker Interchangeability Procedure 2.2.6 operator action to direction to close the 4.16KV #3 -Already Matrix provides location of the nearest 416OVac system Manually Close Circuit Breaker to support installed interchangeable breaker to the one that is E28, sh. 2, Breaker 4160V Circuit associated loads failing to remain closed, whether it is in an Interchangeability Breaker adjacent load center or the warehouse.

Matrix, 4160V The existence and application of this matrix System will be made known to the operators via appropriate training.

269 Operator Action:

[17]

Additional initiation capability #3 -Already This operator action is taken in response to PNPS Procedure Manually Initiate of HPC/RCIC given auto installed provide an alternate high pressure injection EOP-1, RPV HPCI /RCIC initiation signal failure capability during small or medium LOCAs Control Systems and transients. PNPS Procedure EOP-1 assures initiation of those automatic actions important for controlling reactor coolant 52 of 68

Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference inventory. In many cases an operator need only verify system lineups for systems designed to start or operate automatically.

The word "verify" encompasses conditions for which automatic action should have occurred but failed to. In such a case, manual operator action to initiate the appropriate action is required.

270 Proceduralize the

[171 This SAMA would provide the Retain Alternately feeding B1 loads via B3 when 3.M.3-35, Dead operator action to direction to restore B15 and A5 is unavailable can be performed via Bus and Uve Bus feed BI loads via Bi 7 loads upon loss of A5 circuit breaker 52-310. This is an evolution Transfer of 480V B3 When A5 is initiating events as long as controlled by procedure 3.M.3-35. While Load Centers unavailable post-A3 is available. Additionally, there are load restrictions in this trip. Similarly, it would provide the direction configuration as explained in the procedure, feed B2 loads via to restore B14 and B18 loads this is utilized every refueling outage in B4 when A6 is upon loss of A6 initiating support of 4.16kV bus maintenance. This unavailable post events as long as A4 is can restore BI 5 and B137 loads upon loss of trip.

available.

A5 initiating events as long as A3 is available. Likewise, B14 and B18 loads can be restored upon loss of A6 initiating events as long as A4 is available. Modify PNPS procedure to allow this evolution under other circumstances.

271 Provide

[17]

This SAMA would enhance Retain This hard ware modification would provide redundant path the availability and reliability a redundant path for fire water cross tie to from fire of the fire water cross-tie to LPCI loops A and B when either manual protection pump LPCI loops A and B for valve 10-HO-51 1 or 8-1-56 fails to open on discharge to LPCI reactor vessel injection and demand.

loops A and B drywell spray.

crosstie 272 Enhance tornado

[18]

This SAMA would provide

3 -Already Pilgrim IPEEE has verified plant protection PNPS Individual protection for protection from tornadoes installed of equipment require for shutdown following Plant Examination tanks, pumps, and hurricanes the occurrence a tomado. Contribution of for External Events, switchgear, or tornados to CDF is insignificant.

Revision 0, July other equipment/

1994 rooms that may not have I

I 53 of 68

Table RAL5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference protection or that may be susceptible to tornadoes in category F2 273 Structurally

[18]

Failure of the partition wall

3 -Already This masonry wall between the two trains of Calculation Number reinforce masonry between the two trains of Salt installed the Salt Service Water System, designated C15.0.1300 wall in the Intake Service Water results in the as 45.3 is seismically qualified It was drawings C939 Structure loss of both trains, leading reinforced in accordance with a design sheet 1 & 2.

directly to core damage. The developed in response to NRC bulletin 80-improvement would be to 11 "Masonry Wall Design" issued May raise the median seismic 1980.

fragility value from 1.07g to 2.Og by structural beams and columns.

274 Structural

[18]

Structural bracing of the

3 -Already The structural modification to the SBO PNPS Individual modifications to Station Blackout Diesel installed diesel was identified in the PNPS Seismic Plant Examination SBO diesel muffler supports raises the PRA prepared in 1994. The seismically for External Events, seismic fragility to >1.Og induced CDF included the beneficial effects Revision 0, July making this component of this enhancement.

1994 available as an alternate source of emergency power.

275 Structural

[18]

Power from the Station

3 -Already The structural modification to Bus A8 was PNPS Individual modifications to Blackout Diesel passes installed identified in the PNPS Seismic PRA Plant Examination Bus A8 through Bus A8. Minor prepared in 1994. The seismically induced for External Events, structural modifications to the CDF includes the beneficial effects of this Revision 0, July anchorage of Bus A8 enhancement.

1994 increases its fragility to 0.96g thereby providing access to an alternate source of emergency power.

276 Install debris

[18]

Power from the Station

3 -Already This installation of a debris barrier to protect PNPS Individual barrier to protect Blackout Diesel passes installed Bus A8 was identified in the PNPS Seismic Plant Examination Bus A8 through Bus A8. Seismic PRA prepared in 1994. The seismically for External Events, failure of fragile elements on induced CD Frequency includes the Revision 0, July 54 of 68

Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference the main transformer, located beneficial effects of this enhancement.

1994 adjacent to Bus A8, create an interaction hazard that can be removed by the installation of a debris barrier.

277 Seismically

[18]

Seismic restraints on the

3 - Already This enhancement was implemented after Plant Design restrain Nitrogen Nitrogen tanks remove a installed completion of the PNPS Seismic PRA. The Change 97-01 Tanks located seismic interaction hazard seismically induced Core Damage near the thereby raising the fragility of Frequency values do not reflect the Condensate the Condensate Storage beneficial effects of this improvement; Storage Tanks Tanks from 0.1 6g to 0.94g.

however, the impact is estimated at a 2 percent reduction in CDF.

278 Isolate

[18]

This SAMA would limit

3 - Already The hydrogen vent line at Pilgrim has PNPS P&IDs M-combustible combustible source to that installed isolation valves to limit the flammable 226 and M-260 sources for enclosed in line source to that contained in the line.

PNPS Individual seismic or other Plant Examination events for External Events, Revision 0, July 1994 279 Restrain or locate

[18]

This SAMA would eliminate

3 - Already At Pilgrim, flammables cabinets contain ENN-DC-161, flammables probability of cabinets installed small quantities of flammables, usually in Transient cabinets to overturning, spilling the original containers that seal tightly, so Combustible reduce the flammable liquid contents.

overturning a cabinet would not result in Program likelihood of releasing a significant amount of flammable overturning material.

caused by seismic or other events.

280 Ensure that the

[18]

This SAMA would minimize

3 - Already PNPS has a procedure governing the fire-ENN-DC-161, quantity of combustibles and chance of installed safe use and storage of combustible Transient combustible prolonged fire in safety-materials within the process buildings.

Combustible materials in related areas Program critical process areas is monitored 55 of 68

Table RAI.5-1 Phase I SAMA Analysis (SAMA 248 through 281)

Phase I Source SAMA ID Reference Result of Potential Screening Disposition number SAMA Title of SAMA Enhancement Criteria Disposition Reference 281 Monitor and

[18]

This SAMA would reduce fire #3 - Already PNPS Procedure 1.4.3 establishes the ENN-DC-161, control pre-risk installed requirements for the control of site specific Transient staging of outage combustible material storage, ignition Combustible materials sources and impairments of fire systems to Program prevent or minimize the effects of a fire at PNPS. This procedure also provides a control mechanism for tracking system impairments and instituting compensatory measures to minimize the effects that those impairments may have on safety, controls combustible materials within the plant.

56 of 68

Figure RAI.5-1 Fire Water Cross-Tie to LPCI 250 Gallons To LPCI

,000 Fire Water Loops Each Storage Tank A&B T-107B T-1 19 1ZT 1

0H-1 10-CK-570 1

Removable

)riven S p oo rnP 12-P-137 12-T-115 f

12-CK-101 Motor-[

Fire Puf P-135 12-P-136 12-D-7 8-1-21 12-D-3 DieseI-Driven Fire Pump 12-D-1 12-P-132 1.1/2-P-130 To Fire Protection Header 57 of 68

Figure RAI.5-2 PNPS Direct Torus Vent Pathway N1351 OPEN SBGT i7 MINIMUWLOW ORIFICE LOCKED MNli OS TO MAIN STACK EXHAUST PLENUM I--

0z x

a.

TORUS TORUS 58 of 68

NRC RAI 6 Provide the following with regard to the Phase II cost-benefit evaluations:

a. For a number of the Phase II SAMAs listed in ER Table E.2-1, the information provided does not sufficiently describe the associated modifications and what is included in the cost estimate. Provide a more detailed description of the modifications for Phase II SAMAs 3, 6, 7, 10, 20, 21, 22, 27, 28, 29, 35, 43, 47, 53, and 55.

b. Several of the cost estimates provided were drawn from previous SAMA analyses for a dual-unit site (e.g., Peach Bottom). As such, many of those cost estimates reflect the cost for implementation in two units. Since Pilgrim is a single-unit site, some of the cost estimates should be one-half of what has been cited (i.e., Phase II SAMAs 26, 29, 33, 40, 41, 42, 43, 44, and 45) while others are specific to a plant's design, such as the number of valves or batteries that need to be replaced or added (i.e., Phase II SAMAs 38, 46, and 50). For these cases, provide appropriate (specific to Pilgrim) cost estimates.

c. For Phase II SAMA 12, it is stated that probability of vessel failure was modified. Describe the modification considered, and the initial and revised probability of failure.

d. Phase II SAMA 53, control containment venting within a narrow band of pressure, is intended to eliminate failures associated with successful venting. The benefit of this SAMA was determined by reducing the operator failure to vent by a factor of three. It is not clear that reducing the failure to vent probability is related to the actual benefit from this SAMA.

Also, the cost of $300,000 appears high for what appears to be a procedure and training issue. Justify the benefit and cost for this SAMA.

e. In ER Table E.2-1, the percent change in CDF and population dose is reported for each analysis case. However, the change in the offsite economic cost risk (OECR) is not reported. Provide the change in the OECR for each analysis case.

f. Phase II SAMA 47 is stated to include items which reduce the contribution of anticipated transient without scram. Indicate which items are included.

g. Phase II SAMA 49 involves providing instrument signals to open safety/relief valves for medium loss of coolant accident. Discuss whether the signals already exist in the automatic depressurization system.

Response to RAI 6a

a. SAMAs 3 (Install a containment vent large enough to remove ATWS decay heat) and 47 (Install an ATWS sized vent) provide a means to remove decay heat during an ATWS event.

The proposed design modification for these SAMAs involves installation of a larger vent pipe than the existing 8-inch torus vent pipe. The proposed design would require a vent pipe of sufficient size to remove decay heat following an ATWS with MSIV closure and successful recirculation pump and feedwater pump.

SAMAs 6 (Provide modification for flooding the drywell head), and 20 (Provide a method of drywell head flooding), would provide intentional flooding of the upper drywell head such that if high drywell temperatures occurred, the drywell head seal would not fail. The 59 of 68

proposed design modification requires extensive structural modification to accommodate a drywell head flooding system. To flood the drywell head seal at elevation 93-foot, the drywell vent at the 70-foot elevation would have to be plugged and a new penetration would have to be installed in the drywell head at the 93-foot elevation. The new vent penetration would have to be tied into the existing vent line and would have to permit removal of the drywell head at each refueling outage.

SAMAs 7 (Enhance fire protection system and standby gas treatment system hardware and procedure) and 21 (Use alternate method of reactor building spray) would improve fission product scrubbing in severe accidents. The proposed design modification would upgrade the standby gas treatment and fire protection systems to a sufficient capacity to handle postulated loads from severe accidents due to a bypass or breach of the containment.

Loads produced as a result of reactor pressure vessel or containment blowdown would require large filtering capacities.

SAMA 10 (Strengthen primary and secondary containment) would reduce the probability of containment over-pressurization failure. This SAMA is intended for a new plant; hence, it is not practical to back-fit this modification into a plant which is already built and operating.

Since PNPS has a MARK I containment, early release risk is dominated by events that result in early failure of the drywell shell due to direct contact with debris and events that bypass the containment. Strengthening of primary and secondary containment would have a small impact on the overall risk of these accidents. The cost estimated for ABWR was $12 million and the retrofit for an existing containment would cost more. Therefore, the cost of implementation for this SAMA exceeds the revised baseline benefit.

SAMA 22 (Provide a means of flooding the rubble bed) would contain molten core debris on the reactor pedestal and allow the debris to be cooled. The proposed design modification involves a core retention device inside the reactor pedestal area. However, the Industry Degraded Core Rulemaking (IDCOR) Program has investigated core retention devices and concluded, "core retention devices are not effective risk reduction devices for degraded core events". The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, SAMA 22 is not cost effective.

SAMA 27 (Modification for improving DC bus reliability) would increase reliability of AC power and injection capability. It consists of providing an independent DC source, capable of powering both 125 VDC busses. This would be accomplished by feeding the independent source to either DC bus and cross-tying the busses together. This was not found to be cost effective. In response to RAI 5e, the installation of a new DC source to mitigate the loss of DC power initiator is estimated to be $1,953,682 and would increase reliability of DC power and injection capability. This design change was not found to be cost effective. However, the proposed procedural enhancement to cross-tie DC buses as described in SAMA 34 will improve DC bus reliability and is potentially cost beneficial.

SAMAs 28 (Provide 16-hour SBO injection) and 35 (Extended SBO provisions), would improve the capability to cope with longer station blackout scenarios. The proposed design modification for these SAMAs involves adding a battery to improve the coping capability during SBO scenarios.

SAMA 29 (Provide an alternate pump power source) would provide a small, dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power. The proposed design modification would 60 of 68

involve adding one 4.16 KV power source to supply AC power to one feedwater or one condensate pump. The additional diesel generator or gas turbine would have to be sufficiently sized to handle starting (inrush) and running of at least one 5,000 hp pump at a rated voltage of 4.16kV. A generator of that size would easily exceed 6,000 KW which is larger than the existing SBO diesel generator.

SAMA 43 (Improved high pressure systems) would improve prevention of core melt sequences by improving reliability of high pressure capability to remove decay heat. The proposed design modification considers replacing one CRD pump with a flow capacity equal to the RCIC system (400 gpm).

SAMA 53 (Control containment venting within a narrow band of pressure) would establish a narrow pressure control band to prevent rapid containment depressurization when venting is implemented thus avoiding adverse impact on the low pressure ECCS injection systems taking suction from the torus. Hence, the proposed modification for SAMA 53 requires a detailed engineering analysis examining the impact of opening the torus vent path and an examination of the NPSH requirements for LPCI and core spray systems. It would also require an engineering study of the feasibility of closing torus vent valves AO-5042B and AO-5025 against high containment pressures as well as potential hardware modifications.

Procedure changes, simulator changes, and training would also be required.

SAMA 55 (Increase the reliability of SSW and RBCCW pumps) would reduce common cause dependencies from SSW and RBCCW systems and thus reduce plant risk. The proposed design modification would require installation of a different type of pump with a dedicated power supply. In addition, a separate pump intake and a new intake building to house the pump would be required. The proposed dedicated power would be routed from the switchyard underground or overhead to the new intake structure.

Response to RAI 6b

b. Since Pilgrim is a single-unit site, the cost estimates for Phase II SAMAs 26, 29, 33, 40, 41, 42, 43, 44, and 45 are now one-half of what was previously cited (see Table RAI.3-2).

Revision of these cost estimates had no impact on the original conclusions.

Redundant MSIVs are designed to isolate during severe accidents that could lead to radionuclide release and containment bypass. The MSIVs are leak tested each operating cycle to ensure their adequacy. The maintenance rule program monitors the performance of the MSIVs providing early feedback on degradation. In addition, the PSA has determined that the contribution from MSIV isolation failure is insignificant and results in no benefit from implementing this SAMA. Thus, cost estimates for SAMA 38 (Improve MSIV design) are moot.

For SAMA 46 (Increase SRV reseat reliability), the modification assumed replacing 4 ADS/SRV plus 2 RVs with more reliable SRVs. The cost estimate includes engineering analysis and design, and hardware modification. The total cost estimate to implement this SAMA is $1,800,000.

PNPS SRVs have redundant DC power supplies and back up nitrogen supply to enhance SRV reliability. SAMA 50 (Improve SRV design) assumed replacing 4 SRVs with more reliable SRVs. The total cost estimate to implement this SAMA is $1,500,000.

61 of 68

Response to RAI 6c

c. SAMA 12 evaluates a reactor vessel exterior cooling system to potentially cool the molten core before it causes vessel failure, if the lower head could be submerged in water. One method of accomplishing this is to flood containment in accordance with plant procedures during a severe accident. However, the reactor vessel support skirt at PNPS is not vented.

Thus, there would be an air pocket underneath the vessel preventing full contact of the water in containment with the vessel bottom head. Hence, the proposed design modification requires a vent path for the trapped air. This can be accomplished by drilling vent holes in the skirt to provide a vent path for the trapped air or installation of a u-tube extending from the base of the vessel down to an accessible opening in the skirt, then up to a level above that of the containment flood water. The cost of this modification is expected to be similar to that reported for Quad Cities, approximately $2.5 million, and therefore is not considered cost effective.

To account for potential ex-vessel cooling, the containment even tree basic events for vessel failure probabilities were lowered. Event VF_1 represents the probability of vessel failure given core melt less than 20 percent with injection available. VF_1 was changed from 0.05 to 0.025. Event VF_2 represents the probability of vessel failure given core melt greater than 20 percent, core slump and availability of injection. VF_2 was changed from 1.0 to 0.5.

Response to RAI 6d

d. The PSA model assumes the failure of low-pressure injection systems (LPCI and core spray) that take suction from the torus due to inadequate net positive suction head (NPSH) requirements upon performing torus venting. Therefore, the model does not contain basic events for failure of these systems following successful torus venting. Thus, the benefit for SAMA 53 was conservatively estimated by reducing the failure to vent containment basic event. Since CDF is dominated by loss of containment heat removal events, of which failure to vent containment is a dominant contributor, a factor of three reduction in the probability of failure to perform containment venting was considered the appropriate method to evaluate the benefit of SAMA 53. In regards to the cost, as stated in response to 6a, the $300,000 cost includes engineering analyses, procedure changes, simulator changes, and training.

Response to RAI 6e

e. The reduction in the OECR for each analysis case in Table E.2-1 of the ER is given in Table RAI.6-1.

Response to RAI 6f

f. To conservatively assess the benefit of SAMA 047 (Install an ATWS sized vent), the CDF contribution from ATWS sequences associated with containment bypass were eliminated. It is not considered technically feasible to supply enough additional make-up water for injection to allow for large enough removal of ATWS decay heat, beyond the current design basis. Therefore, eliminating the CDF contribution from containment bypass ATWS sequences provides a conservative assessment of the benefits of this SAMA.

62 of 68

Response to RAI 6,q

g. Phase II SAMA 49 provides a means to reduce the consequences of a medium LOCA by increasing SRV reliability to open automatically. This SAMA provides adequate reactor coolant system (RCS) pressure control to prevent an overpressurization condition in the RCS and therefore preclude the occurrence of a LOCA.

The proposed design modification was based on the design implemented at the James A.

Fitzpatrick Nuclear Power Plant called, "SRV Electric Lift System". This plant modification involved opening the SRVs electrically by energizing existing solenoid valves on the pilot stage assembly located on each SRV when the appropriate RCS pressure setpoint is exceeded (the pressures ranges are 1135 psig to 1145 psig). The electric lift initiation is designed to assist the existing mechanical relief in performing its intended function. The SRV electric lift system functions only as an electrical back up to the mechanical setpoint and does not prevent the mechanical portion of the SRV from operating as designed.

Therefore, the proposed design modification does not impact any existing signals in the automatic depressurization system.

63 of 68

Table RAI.6-1 Reduction in Off-site Economic Cost Risk (OECR)

SAMA ID SAMA Description OECR Reduction(%

1 Install an independent method of suppression pool 4.56%

cooling.

2 Install a filtered containment vent to provide fission 20.53%

product scrubbing.

3 Install a containment vent large enough to remove ATWS 1.14%

decay heat.

Create a large concrete crucible with heat removal 4

potential under the base mat to contain molten core 57.79%

debris.

5 Create a water-cooled rubble bed on the pedestal.

57.79%

6 Provide modification for flooding the drywell head.

0.00%

7 Enhance fire protection system and standby gas 1.33%

treatment system hardware and procedures.

8 Create a core melt source reduction system.

57.79%

9 Install a passive containment spray system.

4.56%

10 Strengthen primary and secondary containment.

26.24%

Increase the depth of the concrete basemat or use an 11 alternative concrete material to ensure melt-through 0.57%

does not occur 12 Provide a reactor vessel exterior cooling system 0.19%

13 Construct a building to be connected to primary/

1.33%

secondary containment that is maintained at a vacuum 14 Dedicated Suppression Pool Cooling 4.56%

15 Create a larger volume in containment.

26.24%

16 Increase containment pressure capability (sufficient 26.24%

pressure to withstand severe accidents).

17 Install improved vacuum breakers (redundant valves in 0.00%

each line).

18 Increase the temperature margin for seals.

0.00%

19 Install a filtered vent 20.53%

20 Provide a method of drywell head flooding.

0.00%

21 Use alternate method of reactor building spray.

1.33%

64 of 68

Table RAI.6-1 Reduction in Off-site Economic Cost Risk (OECR)

OECR SAMA ID SAMA Description Reduction (%)

22 Provide a means of flooding the rubble bed.

27.19%

23 Install a reactor cavity flooding system.

57.79%

24 Add ribbing to the containment shell.

26.24%

25 Provide additional DC battery capacity.

2.85%

26 Use fuel cells instead of lead-acid batteries.

2.85%

27 Modification for Improving DC Bus Reliability 1.71%

28 Provide 16-hour SBO injection.

2.85%

29 Provide an alternate pump power source.

5.13%

30 AC Bus Cross-Ties 7.98%

31 Add a dedicated DC power supply.

14.83%

32 Install additional batteries or divisions.

14.83%

33 Install fuel cells.

2.85%

34 DC Cross-Ties 1.71%

35 Extended SBO provisions.

2.85%

36 Locate RHR inside containment.

0.19%

37 Increase frequency of valve leak testing.

0.38%

38 Improve MSIV design.

0.00%

39 Install an independent diesel for the CST makeup 0.00%

pumps.

40 Provide an additional high pressure injection pump with 1.71%

independent diesel.

41 Install independent AC high pressure injection system.

1.71%

42 Install a passive high pressure system.

1.71%

43 Improved high pressure systems 1.14%

44 Install an additional active high pressure system.

1.71%

45 Add a diverse injection system.

1.71%

46 Increase SRV reseat reliability.

0.95%

65 of 68

Table RAI.6-1 Reduction in Off-site Economic Cost Risk (OECR)

SAMA ID SAMA Description OECR Reduction(%

47 Install an ATWS sized vent.

1.14%

48 Diversify explosive valve operation.

0.00%

49 Increase the reliability of SRVs by adding signals to 0.57%

open them automatically.

50 Improve SRV design.

3.04%

51 Provide self-cooled ECCS pump seals.

0.57%

52 Provide digital large break LOCA protection.

0.00%

53 Control containment venting within a narrow band of 2.09%

pressure Install a bypass switch to bypass the low reactor 54 pressure interlocks of LPCI or core spray injection 0.38%

valves.

55 Improve SSW System and RBCCW pump recovery.

7.03%

56 Provide redundant DC power supplies to DTV valves.

3.23%

57 Proceduralize the use of diesel fire pump hydroturbine in 3.04%

the event of EDG A failure or unavailability.

58 Proceduralize the operator action to feed B1 loads via B3 3.23%

58____

when A5 is unavailable post-trip.

Provide redundant path from fire protection pump 18.44%

59_ discharge to LPCI loops A and B cross-tie.

66 of 68

NRC RAI 7 For certain SAMAs considered in the ER, there may be lower-cost alternatives that could achieve much of the risk reduction at a lower cost. In this regard, discuss whether any lower cost alternatives to those Phase II SAMAs considered in the ER, would be viable and potentially cost beneficial. Evaluate the following SAMAs (previously found to be potentially cost-beneficial at other plants), or indicate if the particular SAMA has already been considered. If the latter, indicate whether the SAMA has been implemented or has been determined to not be cost-beneficial at Pilgrim:

a. Use portable generator to extend the coping time in loss of alternating current (ac) power events (to power battery chargers).

b. Enhance dc power availability (provide cables from diesel generators or another source to directly power battery chargers).

c. Provide alternate dc feeds (using a portable generator) to panels supplied only by dc bus.

d. Modify procedures and training to allow operators to cross-tie emergency ac buses under emergency conditions which require operation of critical equipment.

e. Develop guidance/procedures for local, manual control of reactor core isolation cooling following loss of dc power.

f. Enhance loss of salt service water procedure to provide more specific guidance to deal with or prevent a complete loss of the system.

Response to RAI 7a

a. Upon a complete station blackout with failure of the station blackout diesel generator, the 400kw security diesel generator could be used to extend the life of both 125-Vdc batteries.

This allows maintaining RCIC and SRVs availability. Plant procedural changes would be required to implement this SAMA. In addition, since both batteries are required to be able to remove containment heat via the direct torus vent, two cables would be required, one to power each 125VDC battery charger.

With load shedding and battery depletion, core boil off times can reach a maximum of 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months . To assess the impact of prolonging battery life using the security diesel generator to power the battery chargers, the probability of non-recovery of offsite power for 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months was changed to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for SBO scenarios. This resulted in a revised baseline with uncertainty benefit of approximately $212,3623. The estimate cost of implementing and using the portable generator is $75,000. Therefore, this SAMA is potentially cost effective for PNPS.

Response to RAI 7b

b. Loss of battery charging is not a dominant contributor to CDF. This is due to the use of a swing battery charger as a backup to the normally operating chargers. This swing battery charger is powered from MCC B1O which in turn is powered from swing 480VAC load center 3 The value reflects the revised value provided in the response to RAI #3c.

67 of 68

6 which can automatically transfer feed between divisions. The response to RAI 7a discusses use of the security diesel as another method to enhance DC power availability.

Also, Phase II SAMA 34 recommends a procedural enhancement to use DC bus cross-ties to enhance the reliability of the DC power system. Therefore, the proposed SAMA to enhance DC power availability by providing cables from diesel generators or another source to directly power battery chargers has been evaluated for PNPS.

Response to RAI 7c

c. Scenarios involving loss of 125VDC are dominated by failures of circuit breakers to remain shut on main distribution panels. Providing an alternate feed would not mitigate these failures. As discussed in the response to RAI 7b, PNPS uses a swing battery charger as a backup to the normally operating chargers. This swing battery charger is powered from MCC BI 0 which in turn is powered from swing 480VAC load center B6 that can automatically transfer feed between divisions. Therefore, the impact of loss of battery charging is reduced. The response to RAI 7a discusses use of the security diesel as another method to enhance DC power availability. Also, Phase II SAMA 34 recommends a procedural enhancement to use DC bus cross-ties to enhance the reliability of the DC power system. Therefore, the proposed SAMA to provide alternate DC feeds using a portable generator to panels supplied only by DC bus has been evaluated for PNPS.

Response to RAI 7d

d. Phase II SAMA 58 allows operators to cross-tie emergency AC buses under emergency conditions which require operation of critical equipment.

Response to RAI 7e

e. Plant procedure 5.3.26, "RPV Injection during Emergencies," Attachment 1, provides guidance for local manual control of the reactor core isolation cooling (RCIC) system following loss of DC power. Therefore, this SAMA has already been implemented at PNPS.

Response to RAI 7f

f.

Plant procedure 5.3.3, "Loss of all SSW," provides specific guidance to mitigate or prevent a complete loss of the SSW system. Therefore, this SAMA has already been implemented at PNPS.

68 of 68

v t e Letter Sequence Response to RAI
TAC:MC9676, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance Administration Meeting, Meeting Questions 22 May 2006 25 July 2006 23 August 2006 24 October 2006 28 November 2006 Answers 5 July 2006 Results Approval Other: ML061000261, ML061410001, ML063390166, ML070510463, ML072040480, ML072050047
MONTHYEAR2006-07-13 [Table View]

ML061930418

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

Attachment:

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